ML20206E340
| ML20206E340 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 09/19/1997 |
| From: | Howell A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Adensam E NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9905050088 | |
| Download: ML20206E340 (3) | |
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. The licensee Procedure STA-707. and the 10 CFR 50.59 Review Guide were inconsistent in the identification of licensing basis documents (Section 2.1).
The level of detail in certain screenings and evaluations reviewed
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during the inspection was not sufficient to allow an independent review 1
to be conducted without reference to the original change documents (Section 2.2.1).
i One procedure change required by evaluation of a Final Safety Analysis Report revision was not fully implemented and one inconsistency was found between the evaluation and the annual report of evaluations (Section 2.2.1).
Some errors in implementation were noted such as some screenings of design change revisions did not identify the scope of the revision.
One temporary modification was found that did not clearly identify the scope of the temporary modification.
In addition, a couple of " trivial" type dispositions were not properly identified as that type disposition (Section 2.2.2).
Attachments:
Ptesons Contacted and Exit Meeting Attachment ' - Decuments Reviewed
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l DETAILS 1 PLANT STATUS During this inspection period both plants were operating at 100 percent power in Mode 1.
I 2 SAFETY EVALUATION PROGRAM. 10 CFR 50.59 (37001)
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2.1 Proaram The inspectors found that TU Electric had established a good program (STA-707) for reviewing changes, tests, and experiments for changes to the facilities as described by the safety analysis report.
The program for safety evaluations included consideration of other licensing basis documents such as the fire protection report. process control program, offsite dose calculation manual.
q"ality assurance manual. inservice test plan, inservice inspection plan, emergency plan, technical requirements manual, bases sections of Technical Specifications, orders. security plan, safety evaluation reports and supplements, and any commitments in correspondence to the NRC.
Changes to the radioactive waste treatment system were addressed under this program with separate forms (STA-707-3).
The inspectors considered this a complete and extensive definition of licensing basis documents.
The licensee considered the program to conform to the Nuclear Safety Analysis Center (NSAC)-125. " Guidelines for 10 CFR 50.59 Safety Evaluations" which was referenced in the procedure.
The licensee had examined NSAC 125 guidance and had prepared a 10 CFR 50.59 " review guide" for preparers to use in writing safety evaluations. Additionally, the program provided a reference to a design basis document (DBD-ME-28) which established a list of accident events to be considered.
This programmatic assistance was considered by the inspectors to be a positive characteristic of the program.
The inspectors found that the process for evaluations consisted of three steps.
The first step was a screening documented on Form STA-707-1.
For changes. experiments, or tests that could not be resolved by the screening, a second step. called a safety evaluation was performed.
This evaluation was documented on Form STA-707-2. and was used to determine if the proposed activity involved an unreviewed safety question.
If resolution could not be obtained by this evaluation. then the third step. called an unreviewed safety question was taken.
The purpose of the third step was to assure that the change, experiment, or test was not done without prior NRC approval.
The licensee had a training program for preparers and reviewers of screenings and evaluations.
The training consisted of classroom instruction, practical assignments, and an exam.
Requalification training was available but not required. All screenings and evaluations reviewed by the inspectors were prepared and reviewed by a qualified individual.
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The inspectors received inconsistent responses to the question of " changes to what safety drawings constitute a change to the facility?" One engineer replied that only " vital station drawing" changes were to be considered.
This apparent lack of understanding of the licensing basis documents was considered a weakness.
In reviewing program documents the inspectors noted some editorial inconsistencies in the listing of licensing basis documents.
The inspectors noted that 14 different documents were used for the design basis.
Procedure STA-707. Revision 9. listed 11 of these documents in Section 2.1 and 13 others in Section 4.3.
The 10 CFR 50.59 Review Guide listed 12 of these documents.
2.2 Implementation The inspectors reviewed the " Annual 10 CFR 50.59 Report." for 1992.
This report listed 165 evaluations, most of which were design changes.
The inspectors reviewed a sampling of evaluations, arid a sample of su recent operations notification and evaluation (ONE) forms that had evalu? ions associated with them.
2.2.1 Evaluations The inspectors reviewed the 41 evaluations listed in Attachment 2.
The following are the inspectors comments on specific evaluations.
Evaluation 91-112. Clarification of Instrument Air Quality l
This evaluation clarified t he air quality requirements contained in the safety analysis report on air st.: ited to the instrument air system. The licensee maintained their commitment to ensure that the air was free of any 3 articulate greater than one micron. but qualified the description by stating tlat this criteria will nominally be met by using one micron filter having efficiencies of 98 percent or greater. The inspectors reviewed the guidance provided in i
Standard Review Plan. Section 9.3.1. and verified that the licensee referenced l
Industry Standard ISA-S7.3-1975. " Quality Standards for instrument Air."
The evaluation stated that the quality of air supplied at the component instruments will meet the air quality criteria of component manufacturer recommendations.
The inspectors questioned how this evaluation commitment was implemented.
The system engineer stated that Technical Evaluation No. TE 92-1027 verified that instrument air end users are being supplied air that meets the minimum qualitative needs as specified by the manufacturer.
This safety and technical evaluation was considered to be an example of a thorough evaluation and excellent implementation.
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5-l Evaluation 91-146. Structural Heat Sink Calculations This evaluation detailed a revised containment analyses for containment pressure and temperature.
These analyses were performed due to revisions in several input parameters (e.g., containment spray flow reduction, spent fuel pool cooling restoration).
The inspectors reviewed the evaluation and i
determined that it was not thorough and lacked the necessary information to make.a proper assessment of the safety significance of the change.
The inspectors noted that the safety evaluation writeup was inconsistent as presented to the "onsite review committee." The evaluation discussed minor changes to the containment pressure and temperature profiles, yet did not quantify these changes.
Additionally, in at-least two cases, the evaluation i
read "no change in the containment pressure and temperature values." Changes were apparently made to the evaluation at the "onsite safety review committee" meeting characterizing the containmont pressure and temperature variations as "not significant."
l The inspectors further reviewed the referenced calculations and discussed the evaluation with the preparer and licensing personnel.
The evaluation often used simplified " circular logic" in summarizing the impact on the facility.
An example of this was the evaluation section. " Impact on the margin of safety." that did not discuss the quantitative change to the containment parameter profiles. Additionally, the evaluation lacked information that showed that initial containment res)onse was bounded by previous analyses, and that the only impact was later in t1e postulated accident (equipment qualification profiles).
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Also, the evaluation did not adequately discuss that equipment cualification l
was not adversely affected by the two minor instances (one pounc of pressure and three degrees Fahrenheit of temperature) where the analyses' parameters exceeded the bounding curve.
The inspectors were satisfied that the "onsite review committee" understood the evaluation and upon further review of the background calculations, agreed with the licensee that an unreviewed safety cues +. ion was not involved.
This was an example of weaknesses in the level of cetali provided in the evaluation.
Evaluations92-030. Addition of Plant Effluent Holduo and Monitor Tanks. and 92-047. Addition of a Filter Demineralizer Skid to Component Coolina Water System For Unit 2 Both evaluations were performed properly; however, in reviewing the l
evaluations the design change packages were found to have been revised. The revision.; had been subjected to a screening.
Sometimes the screenings could not be related to the scope of the revision.
For example, in Design Modification No.89-168. nine change forms were found with screenings:
however, the screenings did not reflect the scope of the change. for example, the proper document change notices are not referenced.
In Design Modification No.91-061 two screenings were found that exhibited the same problem. These were examples of screenings that did not identify the scope of a revision.
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. Evaluation 92-035. Cold Lea Temoerature Streamina Phenomenon The inspectors found the evaluation writeup to be good; however, the evaluation stated that the Overpower N-16 function was affected by the change.
The change was described as "not significant." The evaluation did not clearly state that the change was in a conservative direction and, therefore. had no adverse impact on safety.
This was an example of an evaluation that lacked detail.
Evaluation 92-081. Final Safety Analysis Report. Section 15.1.4 The inspectors noted that the original screening did not require an evaluation.
However, later review by the licensing department determined that an evaluation was necessary.
The inspectors concurred with the need for an evaluation.
The inspectors observed that this was an example which supported the incorporation of the "10 CFR 50.59 Review Guide" instruction on licensing review of " trivial" type dispositions into a Procedure No. STA-707.
Evaluation 92-089. Condensate Storaae Tank Recirculation Skid The " Annual 10 CFR 50.59 Report" summary for this evaluation states that the piping was installed at an elevation above the minimum required condensate storage tank level required for auxiliary feedwater.
The inspectors reviewed the drawings and walked down the system.
The suction piping to the skid was above the minimum level:
however, the discharge piping from the skid to the demineralized water make-up piping was below the minimum required level for auxiliary feedwater.
The inspectors verified that the evaluation properly evaluated this design and the potential for loss-of-condensate storage tank required minimum level. The new piping connects downstream of the ASME Class 3 seismic boundary check valve (at a Class 5 piping section). Although the annual report was incorrect in its summary of this design change, the evaluation properly addressed the as-built design.
Evaluation 92-185. Uodate Bases For Tank Volumes For Boration Reauirements in Modes 5 and 6 The inspectors found that this evaluation addressed changes in terminology of storage taak definitions in the Technical Specification bases.
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were made to eliminate discrepancies between volumes listed in the bases and I
supporting calculations.
The evaluation was confusing and did not identify the specific values and reason for each change. This was an example of an evaluation that lacked detail.
Evaluation 92-197. Revised Time For Commencina Hot Lea Recirculation This evaluation addressed the impact of increased service water temperature in i
the Final Safety Analysis Report on loss-of-coolant accident analyses.
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evaluation identified the need to revise the hot leg switchover time.
Hot leg recirculation is initiated following a postulated loss-cf-coolant accident to prevent boron plateout in the core.
The change to the hot leg switchover time
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. (from 16 to 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />) was necessary due to the revised post-accident calculation of residual. heat removal subcooled margin. The inspectors reviewed the procedure changes for properly implemented. The body of the text for Emergency Operating Procedure Nos. E0P 1.0A and EOS 1.4A were properly updated. 'However, the checklist attached to E0P 1.0A had not been updated: it still contained the 16-hour requirement.
Licensee persennel committed to update the procedure.
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2.2.2 Screeninos.
The inspectors reviewed the screenings associated with four design modifications, four temporary modifications, and six minor modificatio,s. The screenings are listed in Attachment 2 and are identified by their associated document.
The following are the inspectors comments on specific screenings.
Temocrary Modification 92-2-16. Placement of a Data Acouisition System in the C^ntrol Ro_pm The screening of Revision 1 of this temporary modification was satisfactory; however, it was not clear what the scope of Revision 1 was. The temporary modification, Revision 1. read the same as Revision 0 except for a note on a drawing. The scope of the revision was to land two wires differently not to install an entire system. This was the temporary modification that did not clearly identify the scope of the modification.
Temocrary Modification 92-1-105. Lifted Thermocouole leads of Reactor Vessel Level Instrumentation This temporary. modification permanently lifted a thermocouple lead on one probe in the reactor vessel level instrumentation system.
The probe was clearly defined in the safety analysis report, safety evaluation reports, and Technical Specifications as containing eight sensors.
The change was screened i
as not needing a safety evaluation.. Implementation of the change was clearly a change to the facility as described in licensing basis documents. This temporary r.odification should have been : identified as a " trivial" type change.
-That is, a~ change having "no potential safety impact (e.g., affecting safe shutdown or the safety of o>erations)." This change had little safety significance, since the pro)e was designed to be used with several inoperable
-sensors.
The inspectors discussed with the licensee the necessity for carefully'following their prncedure to preclude changes to the facility that may not be properly evaluated. Additionally, the inspectors discussed the use of the temporary modification process for a permanent change. The systems
- engineer stated that a design modification was used on a similar. sensor that was disabled on another reactor vessel level instrumentation system probe.
The initial approach to this change was to make this a temporary modification.
Later, a similar change was considered permanent. This appeared to be an isolated situation and of minor safety significance.
This was an example of an improperly implemented " trivial" type disposition.
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- - Temocrary Modification 93-1-12. Temocrary Air Source The screening on this temporary modification documented that Question No. 2.
"Did the activity change the facility as described in the licensing basis documents " was answered "no".
Design Modification Nos.93-022 and 93-023, to install permanently similar equipment were screened with a disposition of "yes" to the same question.
The screenings identified that this change was a
" trivial" type change.
Minor Modification No.93-323. Revisina the Vacuum Deaerator Level Transmitter Form One Utilizino a Dry Reference Lea to One Utilizino a Filled and Sealed Reference Lea The inspectors found that the screening on this minor modification identified this to be a " trivial" type change.
That is a change having "no potential safety impact (e.g., affecting safe shutdown or the safety of operations)."
The screening only negatively restated the question and did not detail what supported the conclusion.
2.3 Summary In general, 10 CFR 50.59 screenings and evaluations were satisfactorily performed. A good program has been established with good engineering support of the activities.
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ATTACHMENT 1 1 PERSONNEL CONTACTED 1.1 Licensee Personnel D. Bersi. Mechanical Engineer
- 0. Bhatty, Licensing Engineer
- M. Slevins. Director Nuclear Overview J. Boatwright, Reactor Engineer
- W. Choe, Reactor Engineer
- G. Davis. Systems Engineer C. Feist. Mechanical Engineer
- J. Finneran. Manager Civil Engineering Z. Gajak. Systems Engineer J. Gcode. Technical Staff Training
- T. Hope. P.egulatory Compliance Manager
- J. Kelley, Vice President Engineering and Support
- J. La Marca. Outage Coordinator C. Locke Systems Engineer
- F. Madden. Manager Mechanical Engineering
- G. Merka, Licensing Engineer R. Morrison. Systems Engineering Supervisor
- J. Muffett, Manager of Technical Support / Design Engineering
- D. Pasquale. Temporary Modifications Coordinator C. Poindexter. Systems Engineer L. Pope. Systems Engineer R. Prince. Radiation Protection
- A. Quam Licensing Engineer M. Quick, Systems Engineer R. Smith. Systems Engineering Supervisor M. Smith. Systems Engineering Supervisor
- M. Sunseri, Maintenance Engineering Manager A. Tajoakhsh, Reactor Engineer
- C. Terry. Vice President Nuclear Operations J. Tirrul, Mechanical Engineer
- D. Walling. Manager Electrical / Instrument and Control Engineering
- D. Woodlan. Docket Licensing Manager 1.2 NRC Persor.nel I
- D. Graves, Senior Resident Inspector K. Kennedy, Resident Inspector j
i In addition to the personnel listed above, the inspectors contacted other personnel during this inspection period.
- Denotes personnel that attended the exit meeting.
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-2 2 EXIT MEETING An exit meeting was conducted on September 2. 1993.
During this meeting, the inspectors reviewed the scope and findings of the report.
The licensee did not identify as proprietary any information provided to, or reviewed by, the inspectors.
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ATTACHMENT 2 DOCUMENTS REVIEWED Procedures Station Administration Manual Procedure No. STA-205 " Changes to Procedures."
Revision 16. December 23. 1992 with Procedure Charige Forms 1 through 4 Station Administration Manual Procedure No. STA-602. " Temporary Modifications." Revision 11. August 31. 1993 Station Administration Manual Procedure No. STA-707. "10 CFR 50.59 Reviews."
Revision 9. December 21. 1992 with Procedure Change Form 1 Station Administration Manual Procedure No. STA-716. " Site Modification Process." Revision 9. February 1. 1993 with Procedure Change Forms 1 and 2 "10 CFR 50.59 Review Guide." Revision 1. March 11. 1992 Lesson Plan PTBl.STA.WA1. "STA-707, 10 CFR 50.59 Reviews." Revision 2 Safety Evaluations91-062. Licensing Document Chcnge Request No. SA-92-592 - Radioactive material handling, staging and radioactive waste storage in areas outside the plant.
August 29, 1992 91-074. Licensing Document Change Request - Blank flange at containment hydrogen purge (12"). February 14. 1992 91-089. Licensing Document Change Request - Blank flange at containment isolation valves (48"). July 19. 1991 j
91-096. Licensing Document Change Request No. FP-91-002 - Conduit fire seal (Wisconsin study). August 5. 1991 J
91-105. Design Modification No.90-536 - Temperature monitoring in radiation control cabinets. September 13. 1991 91-112. Licensing Document Change Request No. SA-92-673 - t.larification of instrument air quality. July 15. 1991 91-128. Licensing Document Change Request No. SA-91-162 - Code classification changes in the Final Safety Analysis Report for diesel generator exhaust piping inside the building. February ll. 1992 91-145. Licensing Document Change Request No. SA-91-191 - Nuclear Overview.
December 12. 1991 91-146. Licensing Document Change Request No. SA-90-164 - Containment main steam line break / loss of coolant accident structural heat sink calculations.
February 13. 1992
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o 92-030. Design Modification No.89-168 - Addition of plant effluent holdup and monitor tanks. February 27. 1992 92-031. Temporary Modification No. 92-1-001 - Install adapter with isolation valve on 3/4" X 4" to nipple down stream of ICS-098D to control seat leakage.
January 17. 1992 i
92-035. Nuclear Engineering Manual - 203 Revision 8. Due to cold leg temperature streaming phenomenon. February 17. 1992 92-043. Licensing Document Change Request No. TS-91-019 - Condensate storage tank usable volume. February 25, 1992 92-046.Section XI Program - Cold shutdown testing of component cooling water system Valves PCN-0PT-501A-R2-1. February 27, 1992 92-047, Design Modification No.91-061 - Addition of filter /demineralizer skid to component cooling water system for Unit 2 chemistry. March 9. 1992 92-049. Temporary Modification No. 92-1-007 - Add blank flange to containment ventilation. March 9. 1992 92-053. Licensing Document Change Request No. SA-92-635. Revises Final Safety Analysis Report Table 8.3-2 Sheet 2. 6 and 9 of 11. April 16, 1992 92-056. Licensing Document Change Request No. TR-92-002. Removal of loss of coolant accident analysis credit for the turbine driven auxiliary feedwater pump. March 31, 1992 92-067, Design Modification No.91-059 - Replacement of existing solenoid operated Valves I-HV-8220. 8221 with a different valve design. April 30, 1992 1
92-081. Licensing Document Change Request No. SA-91-108 - Final Safety Analysis Report Section 15.1.4 May 21, 1992 92-084. Licensing Document Change Request - Unit 2 component cooling water to spent fuel pool heat exchanger. May 14. 1992 92-089. Design Modification No.91-033 - Condensate storage tank recirculation skid June 29. 1992 1
92-108. Design Modification No.91-102 - Install a silencer on the auxiliary f
boiler steam supply vent. July 9. 1992 92-116. Design Modification No.92-068 - Modification of the power supplies and controls for nonsafety chilled water recirculation pumps and chiller oil pumps. December 16. 1992
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9 e 92-127. Licensing Document Change Request No. SA-92-697 - Revision of Final Safety Analysis Report tables for diesel generator loading and 125 volts DC loading. November 2. 1992 92-133. Licensing Document Change Request No. SA-91-167 - Filter skid operations. August 7. 1992 92-158. Licensing Document Change Request No. TR-92-017 - Containment spray pump five second delay. November 21. 1992 92-158. Licensing Document Change Request No. SA-93-025 - Revision to safeguards sequencer and response time requirements. February 2. 1993 (Revision 1)92-185. Licensing Document Change Request No. TS-92-036 - Update bases for tank volumes for boration requirements in Mode 5 and 6. November 20, 1992 92-187. Temporary Modification No. 92-2-009 - Load sensing clevis pins (FW).
November 24, 1992 92-188. Temporary Modification No. 92-2-010 - Install lanyard potentiometers and resistance temperature devices at selected locations on the feedwater system piping to assist in measuring piping thermal displacements. December 2.
1992 92-189. Temporary Modification No. 92-2-011 - Load sensing clevis pins (MS).
November 24. 1992 92-195. Temporary Modification No. 92-1-101 - Install tem)orary air to selected components while in Modes 5 and 6 during Unit 1 Refueling Outage 2.
December 1. 1992 92-197, Licensing Document Change Request No. SA-92-811 - Revise time for commencing hot leg recirculation. December 2.1992 92-92-204. Temporary Modification Nos. 92-1-102 and 92-2-016 - Temporary installation of piping vibration monitoring equipment. December 12, 1992 93-021. Design Modification No.93-067 - Bypass uround the spent resin sluice filter. April 21, 1993 93-028. Licensing Document Change Request No. TR-93-010 - Snubber visual j
inspection at two month intervals vice four months. May 3. 1993 93-034. Design Modification No.92-041 - Removal of existing computer Inverter P-2500 computer internals except meteorological and Unit 2 transformer i
equipment. July 12, 1993 93-073. Minor Modification No.93-178 - Provide an additional source of power to Train B radiation monitors. July 2. 1993
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.a.93-084. Temporary Modification No. 93-1-020 - Fuel building crane bypass of the main hoist load cell trip and permissive contacts. August 7. 1993 93-087. Minor Modification No.93-228 - Remove locked valve requirements for several valves in the liquid waste processing system. July 19, 1993 Screeninas Design Modification No.93-022. Addition of trim coolers to cool intake air and cooling water to the instrument air rotary compressors. April 23, 1993 Design Modification No.93-023. Addition of trim coolers for instrument air compressors. May 7, 1993 Design Modification No.93-035. Removal of the underground diesel fuel oil tank. June 8. 1993 Design Modification No.93-036. Revise selected motor operated valves circuits to fulfill the commitment given in NRC Information Notice 92-18. June 21.1993 l
Temporary Modification No. 92-2-016. Place a data acquisition system in the control room. February 26, 1993 Temporary Modification No. 92-1-105. Lift thermocouple leads of reactor vessel level system instrumentation. December 17, 1992 l
Temporary Modification No. 93-2-016. Install feedwater multiplexers in auxiliary building for strain measurements. June 25. 1993 Minor Modification No.93-004. Addition of three floor sleeves with caps to i
l supports future DC power system testing. January 14. 1993 l
l Minor Modification No.93-028. Safety barrier for lower refueling edge. May 6.
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l Minor Modification No.93-032. Revise plant computer system software for l
proper steam enthalpy calculation. May 21. 1993 Minor Modification No.93-044. Setpoint change of diesel generator immersion heaters and temperature alarms. May 21, 1993 Minor Modification No.93-140. Setpoint change on pressure switches. April 28.
1993 Minor Modification No.92-323. Revise vacuum deaerator level transmitter from one utilizing a dry reference leg to one utilizing a filled sealed reference leg. June 15. 1993 l
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