IR 05000213/2003014
| ML17291A864 | |
| Person / Time | |
|---|---|
| Site: | Columbia, Haddam Neck  |
| Issue date: | 06/08/1995 |
| From: | Collins E, Paulk C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML17291A861 | List: |
| References | |
| 50-397-95-03, 50-397-95-3, NUDOCS 9506210058 | |
| Download: ML17291A864 (44) | |
Text
ENCLOSURE 2 U.S.
NUCLEAR REGULATORY COMMISSION
REGION IV
Inspection Report:
50-397/95-03 License:
NPF-21 Licensee:
Washington Public Power Supply System 3000 George Washington Way P.O.
Box 968, MD 1023 Richland, Washington Facility Name:
Washington Nuclear Project, Unit 2 Inspection At:. Richland, Washington Inspection Conducted:
February 13 through March 14, 1995 Team Leader:
C.
P k, Acting Team Leader, Maintenance Branch Division of Reactor Safety Team Members:
P.
Gage.
Reactor Inspector, Engineering Branch Division of Reactor Safety P. Goldberg, Reactor Inspector, Engineering Branch Division of Reactor Safety W. McNeill, Reactor Inspector, Engineering Branch Division of Reactor Safety C. Myers, Reactor Inspector, Engin ering Branch Division of Reactor Safety M. Schlyamberg.
Mechanical Engineer, Consultant 4 -7-95 Date Approved:
mo
.
o
>ns.
c ing ie
.
ngineering rane Division of Reactor Safety a e 95062i0058 9506i6 PDR ADOCK 05000397
EXECUTIVE SUMMARY This team inspection was conducted to assess the overall performance of the
'ngineering organizations at the Washington Nuclear Project, Unit 2 facility, and was performed under the guidance of NRC Inspection Procedures 37550,.
"Engineering."
and 37001,
"10 CFR 50.59 Safety Evaluation Program."
The inspection was performance based, with the team evaluating the performance of the engineers on the basis of the quality of engineering work products (e.g.,
basic design changes, minor modifications.
problem evaluation requests, operability and reportabi lity evaluations.
and calculations)
and overall engineering support of the plant.
Overall. the team found that engineering activities generally were performed in accordance with regulatory requi rements.
The engineering organizations responded well to self-revealing problems; however, the engineering organizations did not exhibit a proactive approach to identify problems.
The performance of'esign engineers.
generally, was in accordance with regulatory requirements.
However.
areas requi ring attention were identified.
These included:
the documentation of assumptions used in calculations, including support for the assumptions; the addressing of uncertainties'n calculations and acceptance criteria; and, the addressing of generic implications for identified problems.
An unresolved item related to the ability of the standby service water system to perform its intended function under design basis conditions.
and a violation with three examples for'the failure to delineate acceptance criteria for inspections and tests were identified.
The team found that management had not been fully effective in identifying and resolving problems.
Engineering management had requested several assessments by quality assurance; however
. corrective actions for the findings were often lacking in addressing generic implications.
Some performance issues identified by the team had also been identified by self-assessment activities but had not been fully addressed.
There were other examples identified by the team where corrective actions were considered to be lacking.
Safety evaluations were found to have been well prepared'n general.
A weakness was identified in the procedures which could result in safety evaluation screenings not being performed.
The performance of technical services (system)
engineers, on the approximately 50 work products reviewed, generally, was also in accordance with regulatory requi rements.
However, weak performance was seen in the resolution of problems and in the relief valve program.
The team noted that while the design engineering and technical services organizations effectively supported the overall plant operation, neither group actively sought out problems.
Both waited for the problems to be self-revealed and then aggressively pursued resolution of that particular problem.
In some cases, the corrective actions were narro The three unresolved items identified in the report will require further review by the NRC.
The first unresolved item, for the ability of the standby service water system to perform its intended function during hot weather months, will require review of a design engineering evaluation.
The second unresolved item, for the relief valve program and status of adjusting rings of various relief valves, will require an inspection of the relief valve program.
And. the third unresolved item, for the acceptability of an operability evaluation for the allowable leakage of the automatic depressurization system check valves. will require review of the basis for a higher leak rate.
111
SUMMARY OF INSPECTION FINDINGS
3.
3'
13
14
19
24
24
25
DETAILS This inspection was conducted pursuant to NRC Inspection Procedures 37550,
"Engineering";
37001,
"10 CFR 50.59 Safety Evaluation Program";
and 92903,
"Followup - Engineering." to assess the overall performance of'he engineering organizations at the Washington Nuclear Project.
Unit. 2.
10 CFR 50.59 SAFETY EVALUATION PROGRAM (37001)
The team reviewed Procedures NOS-47,
"Application of 10CFR50.59 Requirements,"
Revision 2:
PDS-5.
"Design Safety Analysis and 10CFR50.59 Review and Safety Evaluation," Revision 5; and PPM-1.3.43,
"10CFR50.59 Review and Safety Evaluation Process,"
Revision 5, which defined the safety evaluation program.
These procedures defined licensing basis documents as the operating license; the Final Safety Analysis Report: Technical Specifications; safety evaluation reports; the final environmental statement:
the quality assurance program; the emergency plan; the security program and safeguards contingency plan; the fire protection program; the offsite dose calculation manual; and other special reports and documents submitted to the NRC.
The team also reviewed Training Lesson Plan 82-TSR-0400-LP,
"PPM 1.3.43 and PDS-5 10CFR50.59 Review arid Safety Evaluation Process'
" Revision 3.
and found it to be good.
The team found that Section 5.4.4. of Procedure 1.3.43, required the plant operations committee to review safety evaluation screenings for which a safety evaluation was not performed.
However, Procedure 1. 1.5,
"Plant Operations Committee." did not contain requi rements to review safety evaluation screenings.
The team considered this as a weakness.
Licensee representatives stated that it would be addressed.
The team reviewed the following safety evaluations:
SE 94-031, Safety Analysis Change 94-006:
Revise Table 3.9-3
"ASME Class 2 and 3 Active Pumps and Valves," by Removing Check Valve RHR-V-89'ebruary 11, 1994:
SE 94-038, Problem Evaluation Request 294-0043:
RHR-V-8 and RHR-V-9 Leakage, January 23, 26. 28. and'arch 2.
1994; SE 94-040, Plant Modification Record 94-0022-0:
Install Bypass Line Around RHR-V-6A to Allow Controlled Leak-off of Leakage of RHR-V-8/9, February 4, 1994; SE 94-048, Plant Modification Record 93-0251-0:
Remove the Brakes from Standby Gas Treatment Valves'ebruary 12, 1994; SE 94-049, Plant Procedure Manual 8.3.318.
Revision 0:
Add Differential Pressure Testing of SW-V-187A and SW-V-188A. February 12, 1994; SE 94-103. Safety Analysis Change 94-033:
Delete Stroke Time and Differential Pressure Requirements for RCIC-V-10 ~ February 24, 1994;
~
SE 94-117, Field Change Request 93-0157-0-02:
Deviates the Mounting of Four Solenoids to a Near Location.
March 27.
1994:
~
SE 94-136.
Basic Design Change 94-0152-OA:
Motor Replacement on RCIC-MO-63. June 1.
1994;
~
SE 94-193, Field Change Request 83-0107-2-21:
Re-evaluation of Fuel Oil Transfer
. September 8.
1994;
~
SE 94-195.
Instrument Setpoint Change Notice 1172:
Increase Migh Temperature Alarm, July 22.
1994;
~
SE 94-212.
Licensee Controlled Specification
~ 7.2. 1:
Add'ition of Control Room Emergency Chiller Requirements.
October 27, 1994; and, SE 94-225, Relief Request 2IST-12:
Exception to the 1989 Code with Regard to Defect Removal, December 6,
1994.
1. 1 Safet Evaluation SE 94-040 Safety Evaluation SE 94-040 was prepared to support a modification to install a bypass valve around residual heat removal system Valve RHR-V-6A.
The team found that Amendment 49 to the safety analysis report dated August 1994 changed the text to reflect the installation of the bypass line; however, the safety analysis report drawings'eneral Electric Piping and Instrument Drawings 02E12-04, 10,1; Sheets 1 and 2. were not changed to reflect the installation of this modification.
Further, the team found that the safety analysis report drawings.
required to be changed as a result of this modification. were not identified in the design change package.
Procedure EI 2.8.
"Generating Facility Design Change Process,"
Revision 11, Section 4. 1.8.a.
stated that ".
.
. [drawing change noticesj DCNs shall be prepared in accordance with Attachment 5. 10."
Through this process, the preparer would indicate the changes to be made on a given drawing, including safety analysis report drawings.
In Attachment 5. 10. the preparer was requi red to list all drawing change notices for the design change on the document control system input sheet.
As of March 3, 1995, the team found that Drawing 02E12-04,10.
1 was not identified on a drawing change notice, nor on a
document control system input sheet.
The fai lure to identify Drawing 02E12-04, 10. 1; Sheets 1 and 2 as a drawing requiring revision on the document control system input sheets was identified as an example of'ailure to implement a procedure (397/9514-01).
In response to the team identifying this violation.
a Request for Technical Services 95-03-001 was prepared by engineering personnel to revise the drawings in question.
This violation was cited because the team did not consider the issuance of a request for technical services to have been'comprehensive corrective action to prevent recurrenc.2 Safet Evaluation SE 94-193 Safety Evaluation SE 94-193 was prepared to evaluate the mixing of auxiliary boiler and emergency diesel fuel oil as a result of a modification to install a filter polisher in the diesel fuel oil system.
In review of this safety evaluation.
the team found that two earlier safety evaluation screenings had been performed for the modi'fication on February 20, 1991, and January 23.
1992. but concluded that no safety evaluations would be required.
Neither of these screenings identified a change to the fire hazards analysis.
Procedure PDS-5.
"Design Safety Analysis and 10CFR50.59 Review and Safety Evaluation,"
Revision 2, Attachment 7. 1,
"10CFR50.59 Review Guidance,"
required that fires, their effects.
or affects on the fire hazard analysis.
were to be considered for the performance of a safety evaluation.
This modification resulted in a change to the facility, as described in the "Fire Protection Evaluation,"
which was part of the safety analysis report, and affected the analysis of fire hazards.
The failure to perform a safety evaluation evaluating the fire hazard was identified as another example of failure to implement a procedure (397/9503-01).
Again, this example was of low safety significance on its own, but was considered in aggregate with the other example to be indicative of an area needing licensee management attention.
1.3 Conclusions The team concluded that the safety evaluations, in general.
were well prepared and documented; only a few evaluations could not stand alone.
A weakness was identified.in Procedure 1. 1.5 which could result in safety evaluation screenings not being reviewed.
The team also identified two examples of a violation for failure to implement procedures.
ENGINEERING (37550)
2.1 Desi n
En ineerin 2.1.1 Design Hodi fications The team reviewed modification packages to verify that the packages were complete and accurate.
The review included verification that the description of the modification, the safety evaluations, installation instructions, and post-modification testing requirements were adequate.
In some cases, other supporting records associated with the modifications, such as calculations and other engineering documents were selected and reviewed to verify the adequacy and accuracy of the engineering process.
2. 1.1. 1 Modification PHR-02-86-0324-0 This modification was implemented to prevent a water hammer event during the alignment of a backup water supply for the spent fuel pool cooling system from the standby service water system.
This was accomplished by installing limiting flow orifices.
The team found that operating Procedure 2.8.5,
"Fuel Pool Cooling and Cleanup System,"
Revision 19, requi red Valve SW-V-188A(B),
the standby service water alternate cooling supply motor-operated valve to the spent fuel pool cooling system heat exchanger'o be manually opened to provide the flow path.
The procedure also contained a note that stated that the valve should be approximately 14 percent open when the desired flow was obtained.
As a result of the addition of 'the orifice, the team considered that the valve would need to be more than 14 percent open.
However, the team found that Procedure 2.8.5 was not identified as needing to be revised as a result of this modification.
The team considered that the procedure could be misleading from a human performance stand point.
Licensee representatives acknowledged this discrepancy and stated, prior to the end of the inspection, that this procedure would be considered for revision.
2. 1. 1.2 Minor Design Change MOC 94-0194-0A The team reviewed this minor design change which was implemented to remove Valve MS-V-152A, a nonsafety-related motor-operated valve for the blow-off drain on a main steam bypass valve strainer.
This valve was used for plant startup and was not required for normal plant operations.
Since the valve developed a leak and could not be repaired'his modification was prepared to delete the valve.
However, this modification was limited to removal of the valve only and did not address removal of the wiring for the motor-operator, including switch and position indicators located in control room.
The review of the impact of declaring the wiring to be spared, which did not include a'uman factors review, was deferred.
This was of concern to the team because the switch was left installed on the control board without any indication that it was non-functional.
The team noted that.
although this modification included some electrical work, the preparation, review, approval and safety evaluation screening were performed by mechanical engineers only.
There was no documented evidence that electrical engineers reviewed this modification.
Although this modification was nonsafety-related.
the same procedure and process was used for the safety-related modifications.
The absence of electrical engineers'nvolvement was identified as an example of a lack of rigor in that the emphasis was to remove the physical valve quickly, not to implement a thorough modification.
Additionally, the human factors aspects of the modification were deferred.
2. 1. 1.3 Basic Design Change BDC 92-0178-6G The team reviewed Basic Design Change BDC 92-0178-6G which was implemented to change the design and the operating pressure of the diesel starting air system for the Division III (high pressure core spray) diesel generator.
The team found that Valves DSA-SPV-5C1 and DSA-SPV-5C2. the starting air system solenoid valves.
were requalified by licensee personnel to function at the maximum system operating pressure of 1896 kPa (275 psig)
and not the new system design pressure of 2241 kPa (325 psig).
The justification for this new pressure rating.
2241 kPa.
was that the bodies of the solenoid valves were designed for pressure in excess of 2241 kPa; that operation of the solenoid valves at pressures greater than the design pressure would requi re multiple
component failures in the air start system.
since the diesel starting air system pressure was controlled by Class 1E pressure switches; and, that the maximum set point value of the switches, including instrument uncertainty, was less than 1896 kPa.
The team's review of the set point data sheets for the pressure switches indicated that there was no reference to the solenoid valve qualification limits.
Lacking such information, there was a potential for a change of the pressure switch setting to a higher value.
Licensee representatives acknowledged this concern and stated that they would revise the set point data sheets to provide a cross reference to the qualification of the solenoid valves.
A number of pages (i.e.,
BOC 92-01786G-72.
-324 through-341) were prepared by the same engineer who also provided the cognizant engineer
"approval."
The team was informed that this was a
common practice and that the "approval," in reality. signified that the package was completed and ready for verification.
According to Procedure EOP 1. 13.
"WNP-2 Engineering Technical and Hanagerial Responsibilities,"
Revision 3, only supervisors and above can approve design work.
The approval of the modification by the same individual who prepared portions of the package was not in compliance with this procedure, as well as being contrary to the requirements of ANSI N45.2. 11.
Licensee representatives acknowledged this concern and stated that they would review their procedure for potential need of additional clarifications.
The team concluded that the use of the term "approval" was an example of a weakness in a procedure because the exact purpose of the signature was not stated.
2. 1. 1.4 Hinor Design Change 90-0266-OA This modification was implemented to remove Valve RV-2.
a thermal relief valve, from the tube side of the nonregenerative reactor water cleanup heat exchanger.
The team found that design Specification 23A1878.
Item 4.3.3. 14, stated that "Lt]hermal relief valves on the shell and the tube side of the regenerative heat exchanger and on the tube side of the nonregenerative heat exchanger shall be provided
.
.."
Although the design'ngineers provided a
justification for use of only one thermal relief valve for the tube side of the regenerative and the nonregenerative heat exchangers, they failed to address why this modification was acceptable since it, was not in agreement with the design specification.
The team considered that the minor design change met code requirements; however, the team found no objective evidence that the requirements stated in this specification were considered in the preparation and implementation of this minor design change.
The team considered this to be an example of lack of rigor on the part of design engineers in that consideration was not given to the revision of the design specification.
2. 1.2 Calculations The team reviewed 17 calculations that were performed in support of modifications or had been revised during the 12 months preceding the inspection.
The team reviewed the administrative procedure for the preparation of calculations.
The team found that the guidelines were weak in
that they did not provide comprehensive instructions to address all aspects of the design basis.
The team noted that calculations performed for Technical Specification setpoints contained a margin for instrument uncertainties.
2. 1.2. 1 Standby Service Water System The team reviewed Calculation ME-02-93-004, Revision 0, for the standby service water system which addressed how diversion of the standby service water for the spent fuel pool cooling affected standby service water system performance.
The team noted that the standby service water pumps did not have any flow requirements imposed by Technical Specifications.
The licensee relied on an annual standby service water flow balance for the operability determination of the standby service water system, and the test did not provide a flow path for the spent fuel pool cooling system.
The team found that the model used in this calculation was not validated,. thus. the uncertainty of the predictions of this calculation was not addressed.
The team also noted that the calculation did not account for the spray Dond level loss during the 30 days following a design basis accident.
Although this calculation used actual recorded data't did not account for the measurement uncertainty.
The acceptance criteria did not include any correction for pump degradation, measurement uncertainties, or variations in spray pond 'level.
'he team noted that the test results indicated sufficient flow was available.
The acceptance criteria for in-service pump testing for the standby service water pumps was based on a curve.
The pump degradation limit was not a fixed reference point, but rather the test acceptance was based on the degraded pump curve plotted with a uniform 10 percent head reduction.
A minimum (system performance based)
design limit for flow was not established consistent with the design calculations.
Therefore,'the pump performance could be within the area bounded by the pump curve and the in-service testing degraded curve, but the flow could be less than requi red for the design basis accident.
Criterion III of Appendix B to 10 CFR Part 50 states, in part, that "design control measures shall be applied to
.
.
. delineation of acceptance criteria for inspections and tests."
The failure to delineate acceptance criteria to demonstrate the ability to perform safety functions under design basis conditions by not including pump degradation and variations in spray pond level in Procedures 7.4.7. l.l. 1 and 7.4.7. 1. 1.2 for the demonstration of system operability as of March 3, 1995, was identified as an example of a design control violation (397/9503-02).
Another major input to Calculation ME-02-93-004 was the standby service water flow requirement for the resi'dual heat removal heat exchanger.
This flow rate was derived in Calculation ME-02-93-05, using a tube side fouling value of 0.002.
The team reviewed the residual heat removal heat exchanger test records to determine if the as-measured values were comparable with the design assumption The performance monitoring tests were used by design engineers to determine the ability of the residual heat removal heat exchangers to remove the required heat loads during a design basis accident.
These tests were also used to collect data for the determination of the fouling factors.
. The fouling factors, calculated on the basis of the results of residual heat removal heat exchanger testing, varied from 0.0009 to 0.00143.
However, the test data used for fouling factor calculations were not corrected for the instrument uncertainty.
Industry experience has indicated that instrument uncertainty alone.
expressed in terms of fouling. can be as large as the above factors.
Therefore, with instrument uncerta'inties, the actual fouling factor could have been greater than the 0.002 assumed in the calculation and could have resulted in less heat transfer than needed because Calculation HE-02-93-05 may not be sufficiently conservative.
The licensee considered that there was no short-term operability concern since the standby service water temperature, during the inspection, was significantly lower than the design basis temperature.
Additionally, engineering personnel performed a preliminary review to determine if the requi red standby service water flows could be provided unti 1 the unit outage, scheduled for April 1995.
This review considered maximum design basis standby service water temperature between October and April.
The engineers concluded that some margin existed between the predicted and required flows and that this margin was sufficient to deal with the standby service water uncertainties discussed in this report.
Furthermore.
the performance of the standby service water was routinely monitored, providing the engineers with confidence that the plant could be safely operated until the outage.
The team concurred with the engineers'ssessment of the operability of the standby
. service water through April 1995.
The licensee stated that they intend to address all standby service water iss'ues before the end of the outage.
The review of the evaluation of the standby service water system's ability to provide adequate flow to the residual heat removal heat exchangers with fouling factors adjusted for instrument uncertainties.
and the review of the service water system requirements for hot weather months were identified as an unresolved item (397/9503-03).
2. 1.2.2 Diesel Fuel Oil System The team reviewed the surveillance test procedures for the diesel fuel oil transfer pumps and found that the acceptance criteria were not cor rected for changes in suction pressure.
level. temperature, and density.
Also, the team found that the criteria did not account for pump degradation.
Review of this issue by design. engineers indicated that there was no operability concern since current and past diesel fuel oil system performance flow had sufficient margin to accommodate these variables.
The design engineers demonstrated.
to the team, that the uncertainties were less than the
difference between the test results and the acceptance criteria in this case.
However. if the actual flows had been allowed to drop to the level derived in the calculation, there may not have been sufficient flow to maintain the emergency diesels operating under design basis conditions.
Criterion III of Appendix 8 to 10 CFR Part 50 states, in part, that "design-control measures shall be applied to
.
.
. delineation of acceptance criteria for inspections and tests."
The failure to delineate acceptance criteria to demonstrate the ability to perform safety functions under design basis conditions by not including pump degradation or variations in suction pressure.
temperature.
or density, in the surveillance test procedure for the diesel fuel oil transfer pump for the demonstration of system operability as of March 3.
1995.
was identified as another example of a design control violation (397/9503-02).
2. 1.2.3 Spent Fuel Pool Cooling System The team reviewed Calculation 5.35. 14. Revision 3, which established the design basis for the spent fuel pool cooling system.
The team found that this calculation did not address a runout condition (i.e., operating at maximum flow with minimum discharge pressure).
The design engineers performed the available net positive suction head calculation for a selected range of flows'sing a maximum flow of 2271 Lpm and concluded that the system was operable.
The team's review of this calculation found that at 2366 Lpm (625 gpm). the extrapolated value for the available net positive suction head would be close to the requi red net positive suction head.
The pump head curve= and the required net positive suction head curve used in this calculation were generic and not certified.
Thus, it was difficult to draw a conclusion on the maximum flow value and the available net posjtive suction head adequacy.
The licensee concluded that there was no condition adverse to safety, since the spent fuel pool cooling system had a backup safety-related makeup source.
thus a potential failure of the spent fuel pool cooling pumps would not result in the loss of the spent fuel pool cooling function.
The team reviewed the test procedures for the spent fuel pool cooling pumps and found that neither the test results.
nor the acceptance criteria, accounted for changes in suction pressure (e.g.,
surge tank level) or measurement uncertainties.
The design engineers did not identify a calculation to demonstrate that the maximum change in the surge tank level would not result in a reduction of flow below the minimum acceptable of 2271 Lpm (600 gpm), nor below the design required flow of 2177 Lpm (575 gpm).
The team could not connect the actual acceptance criteria with any specific system design documents.
The team concluded that the acceptance criteria was derived from the design specification for the spent fuel pool cooling heat exchangers.
The design engineers concluded that there was no condition adverse to safety, since current and past spent fuel pool cooling system flow performance was in excess of the ASME Code Section XI limit. and the uncertainties were less than
the difference between the test results and the acceptance criteria.
However, the engineers could not demonstrate that. if the flow was reduced to the level derived in the calculation.
the system would have been capable of performing its design'basis function.
Criterion III of Appendix B to 10 CFR Part 50 states.
in part. that "design control measures shall be applied to
.
.
. delineation of acceptance criteria for inspections and tests."
The fai lure to delineate acceptance criteria to demonstrate the ability to perform safety functions under design basis conditions by not including pump degradation or variations in suction pressure spent fuel pool cooling test procedures for the demonstration of system operability as of March 3.
1995.
was identified as another example of a design control violation (397/9503-02).
2. 1.2.4 Calculation ME-02-94-42. Revision
This calculation determined how long the emergency diesel generators could operate following a loss of the standby service water system under a specific Appendix R scenario.
The team noted that this calculation did not address the available net positive suction head limitation for the jacket water pumps which could contribute to the potential loss'r significant reduction, of jacket water cooling.
After the team identified this issue, the cognizant
'ngineer reduced the maximum allowable (calculated)
jacket water temperature from a near saturation value of 110. 1'C (231.'F) to 97.8'C (208'F) which led to a reduction of the maximum available time for the operator intervention from 9. 17 minutes to 7 minutes.
The licensee concluded that this time was acceptable.
The team also f'ound that the licensee did not have the vendor 's concurrence for a calculation assumption.
The licensee subsequently obtained the concurrence.
The fai lure to have documentation supporting assumptions was considered a lack of rigor.
2. 1.2.5 Heat Exchanger Tube Plugging Limit The team found, through discussions with a design engineer, that there were no formally established tube plugging limits for safety-related heat exchangers; however, calculations for the safety-related heat exchangers (e.g.,
Calculation ME-02-93-05)
assumed a tube plugging limit of five percent.
The team was informed that none of the heat exchangers exceeded this limit.
This lack of the linkage between the design assumptions and the plant operating practices was of concern to the team.
because it illustrated a potential for configuration control breakdown and a lack of rigor in the performance of calculations.
2. 1.2.6 Motor-Operated Valve Design Basis Review for Feedwater Valves The team reviewed Calculation ME-02-94-22. Revision 0, which was performed to determine the maximum expected differential pressure for feedwater isolation valves and to determine torque switch requirements.
The team found that instrument uncertainty was not accounted for in determination of the maximum
expected differential pressure or torque switch set points.
In addition to the lack of instrument uncertainties, the engineer did not utilize adequate assumptions to determine the maximum expected differential pressure.
The subject valves had been added to the motor-operated valve program on the basis of a probabi listic risk assessment, even though they were nonsafety-related.
The team found that the valves were being reconsidered for exclusion from the motor -operated valve program on the basis of the results of this calculation.
The team was informed that this was the only calculation that had been performed by engineering personnel to determine switch settings for motor-operated valves; the others had been performed by a contractor.
In this
.
calculation, the engineer used a differential pressure calculated from the pump discharge pressure and reactor vessel pressure.
The team noted that the engineer did not consider a feedwater line break to be the worst case for differential pressure.
This was a concern to the team because it demonstrated a lack of understanding of the parameters used to select valves and 'determine setpoints.
2. 1.2.7 Calculation. ME-02-93-09, Revision
This calculation was performed to determine the effects of the addition of'he control room chiller on the ultimate heat sink.
After this calculation was issued, another Calculation.
ME-02-92-41, was revised to address the impact of the control room chiller load on the ultimate heat sink.
The team noted that Calculation ME-02-92-41 used a greater control room chiller heat load value than Calculation ME-02-93-09.
The cognizant engineer acknowledged this discrepancy and stated that Calculation ME-02-92-41 was the calculation of
'ecord and that the status of Calculation ME-02-93-09 would be changed to historical.
2. 1.2.8 Conclusions The team found problems in a number of calculations.
These problems included the basis for assumptions.
completeness of assumptions.
and the addressing of instrument uncertainties.
Some of these problems were identified by the team as violations.
These problems demonstrated a lack of'igor in the performance of calculations.
2. 1.3 Design Engineering Problem Evaluation Requests 2. 1.3. 1 Problem Evaluation Request 294-0039 The fai lure to comply with the procedural requirements for documenting the qualifications of design engineers was identified in Problem Evaluation Request 294-0039.
which was initiated by a quality assurance auditor on January 14, 1994.
The team found that the dispositioner had adequately addressed the concern for the design engineers'ut had not addressed other engineering organizations in response to this request.
The team considered this response to be an example of not addressing generic implications.
This problem evaluation request was closed on August 17, 1994.
Engineers provided the team with five additional problem evaluation requests for what were considered to be qualification documentation concerns.
These additional documents were presented to indicate that generic implications had been addressed.
On November 10, 1993, Problem Evaluation Request 293-1318 was initiated to
.
identify the deferral of corrective actions for qualification deficiencies related to operators, identified by the training department in October 1992.
The cause attributed to the deferrals was "poor planning and inadequate resources."
The initial corrective actions were:
to stop using the personnel qualification database for assigning work; to verify the accuracy of the qualification matrices; and, to initiate a deviation to the procedure to remove the reference to the personnel qualification database.
The problem evaluation request was closed on April 8, 1994.
On November 16, 1993.
Problem Evaluation Request 293-1341 was initiated as the result of a quality assurance audit of training and qualification of the plant staff.
The discrepancies included a lack of training matrices.
incomplete records of training and development plans.
and insufficient records in the Technical Services Division training files.
The corrective actions were:
to establish an engineering support staff training advisory group; to implement a
new program description.
TTH-5.2.4; to identify and formally train coordinators in program requirements'nd to discuss this problem evaluation request and the lessons at the next engineering support staff training advisory group meeting.
This problem evaluation request was closed on September 1994.
On November 24, 1993, Problem Evaluation Request 293-1359 was initiated as the result of a quality assurance audit during which the auditor identified a lack of qualification documentation for a radiological waste vendor and two health physics technicians.
The corrective action for this finding was to "develop a
checkin/checkout system for new employees and contractors and vendors to remind supervisory personnel to assure administrative requirements
.
.
.
[werej adhered to."
This problem evaluation request was closed on April 28.
1994.
On June 10, 1994, Problem Evaluation Request 294-0551 was initiated by a technical specialist to identify that a contract engineer and a health physics supervisor were not qualified by the responsible department manager.
The probable cause was "(i]nadequate/ineffective corrective action from PER 293-1359."
The corrective actions were:
to perform qualification in accordance with plant procedures for the two people; to perform a company wide se'arch for any other contractor personnel who should be qualified in accordance with plant procedures, but were not:
and, to heighten the sensitivity of plant management to the requi rements of the plant procedures.
This problem evaluation request was closed on August 15, 1994.
On August 22, 1994 'roblem Evaluation Request 294-0803 was initiated by an engineer to identify several departments that did not have a program for qualification of personnel, or documentation of a program, in accordance with plant procedures.
The probable cause was documented as "Lljack of awareness of requirements."
The immediate corrective action was to discuss the issue
with the applicable department management per sonnel.
The recommended corrective actions were to identify other departments that should have programs and to develop a program to ensure that procedural requirements were implemented.
This problem evaluation request was closed on October 4.
1994.
The team found that each of the problem evaluation requests were resolved for the individual issue identified. but no review of qualification documentation requirements.
programs, or records for the entire site was performed.
Each problem evaluation request had been closed, but three of four qualification documents reviewed by the team during the inspection were not complete.
The qualification documents were for "qualified" system engineers; however, the Manager, Technical Services Division, or his designee, had not signed the forms indicating qualification as system engineer.
The team noted that Section 6.3. 1 of Procedure 1.3. 12A, "Processing of Problem Evaluation Requests (PER)," Revision 0, required the dispositioner to document the "action taken or to be taken
.
.
. to correct the problem and to prevent recurrence
.
.."
The team also noted in Attachment 8.2,Section II.B.6. that the dispositioner was to "[p]rovide an assessment of the generic impact of the condition."
The team concluded that the evaluation of Problem Evaluation Request 294-0039 was weak in that the engineers (dispositioners)
did not address the generic implications of failure to document the qualifications of design engineers by confining the resolution only to the Engineering Directorate.
Also, the engineers did not document corrective actions that would prevent recurrence of the problem.
2. 1.3.2
. Problem Evaluation Report 294-0527 On June
~
1994'n engineer identified that the environmental qualification data package (QID 036005) for Brand-Rex coaxial cable had been voided in October 1990.
The licensee continued plant operations with this cable installed in the plant without a current qualification data package as required by 10 CFR 50.49(d).
The team noted that. after identification in 1994. engineering personnel promptly reissued the qualification data package, thus complying with the regulatory requirements.
The team also noted that the manner in which qualification data packages were maintained had been modified, requi ring data packages for any component that may be installed in the plant.
In accordance with the guidance of 10 CFR Part 2, Appendix C, Section VII.B.(2). this violation was not cited.
2.1.3. 3 Conclusions The team concluded that the engineers'nvolvement in plant problems associated with motor-operated valves was timely and technically thorough.
However, the generic implications of problems associated with design engineer qualifications were not adequately addressed.
2.1.4 Molded-Case Circuit Breaker Testing The team reviewed the design controls used by design engineers to establish the instantaneous trip setpoint for molded case circuit breakers supplying power'o motor-operated valves.
The team reviewed Procedure EES-5,
"General Fuse Selection Criteria and the Electrical Protection of 460 VAC and 125-250 VDC Motors." Revision 2.
The team reviewed the surveillance test procedure for periodic testing of containment penetration protection ci rcuit breakers as required by Technical Specification 4.8.4.2.
The team reviewed Procedure 7.4.8.4.2.3,
"Penetration Conductor Low Voltage Circuit Breaker," Revision 4.
In addition. the team reviewed preventive maintenance Procedure 10.25.48.
"Testing Molded Case Circuit Breakers,"
Revision 6.
The team found that preventive maintenance testing was scheduled to be performed every 4 years for molded-case circuit breakers.
The team noted that 120 VAC breakers were not included in the preventive maintenance testing.
In response to the team's observation.
the engineers identified that no 120 VAC circuits inside containment.
protected by circuit breakers, were currently energized during plant operation.
A technical evaluation request was initiated to review the adequacy of the testing of 120 VAC circuit breakers inside containment.
The team noted that Licensee Event Report 93-14 had identified previous deficiencies in the administrative controls which had been established to deenergize 120 VAC circuits inside containment.
The team found the action for this event report to be adequate.
'The team noted that Procedure 10.25.48 required as-found trip current data.
However, the team found that maintenance personnel performed a cleaning and lubrication as well as manually exercising the breakers prior to over-current trip testing during testing of molded-case ci rcuit breakers
.
While some alteration of the as-found condition of the breaker was unavoidable to facilitate testing, the team questioned whether the intentional cleaning, lubricating, and exercising of the breaker prior to trip testing was appropriate since the trip testing was intended to identify the as-found condition of the breakers.
The team was concerned because the exercising was performed solely to improve breaker performance, and test data obtained after exercising may not be representative of the as-found performance of the breaker.
In response to.the team's concern, design engineers initiated a problem evaluation request to review the adequacy of their molded-case circuit 'breaker test procedures.
The engineers stated that the breaker test procedure was currently being updated and that additional enhancements to address the team's concerns would be evaluated.
The team found the proposed actions to be adequate.
2.2 Technical Services S stem En ineerin 2.2. 1 Problem Evaluation Requests The team reviewed approximately 50 problem evaluation requests that had been responded to by technical services (system)
engineers during 1994.
The team had the following comments
'on problem evaluation requests assigned to system engineers.
2.2. 1. 1 Problem Evaluation Requests 294-0116 and 294-0178 These requests were initiated upon the identification of fuse discrepancies between the as-found installation and the design drawings.
The team found that a fuse control program had been in effect since approximately 1983.
This program was intended to result in the identification and correction of discrepancies such as these.
with an estimated completion date of December 1996.
While no safety-significant problems had been identified, the team considered that the length of time to complete the fuse identification program was excessive.
2.2. 1.2 Problem Evaluation Request 294-0480 During an attempt to provide power to the plant by backfeeding through the main transformer on May 24, 1994, the lockout relays actuated, opening the 500 kV breakers.
The dispositioner determined that the cause was the result of jumpers.installed on the main transformer current transformers for the performance of another test on the main generator.
The dispositioner documented that "[t]his event has no generic impact as it is [sic] was caused by the uncommon event that these two procedures would be worked together."
The corrective actions were to revise the affected procedures to prohibit the backfeed operation while testing of the main generator was in progress.
The team found that the control room operators had inquired into the status of the testing.
The operators were informed that the test had been completed.
The operator did not ask if the system had been returned to normal, nor did the technician provide-that information.
The team concluded that a
contributing cause was a communication breakdown, which was not addressed by the dispositioner in response to this evaluation request.
Additionally, the team reviewed the actions taken in response to a similar event that occurred in 1990 to ascertain if the actions taken should have.
or could have, prevented the 1994 event.
The team questioned the system engineers about an event that occurred during the 1990 refueling outage.
During that outage, plant personnel were attempting to backfeed power and damaged the potential transformers because grounding jumpers installed for testing had not been removed.
The evaluation of that event resulted in changes to the backfeed procedure, but was not thorough enough to have
identified any other effect of grounding or shorting jumpers remaining installed on components associated with the backfeed operation.
The team was informed by the system engineers that the problem which resulted in the 1994 event would not have been identified other than by its occurrence.
The team concluded that the failure to address the previous event was another example of weakness in implementing Procedure 1.3. 12A.(Section 2. 1.3. 1).
2.2. 1.3 Problem Evaluation Request 294-1089 This problem evaluation request was prepared to address a drawing error identified by control room personnel.
The control room personnel noted that pressure switches for the diesel starting air system were incorrectly identified on Drawing EWD-47E-027, "Electrical Wiring Diagram Standby AC.Power System Diesel Generator 1 Diesel Start Air Compressors DSA-C-1Al and DSA-C-1A2," Revision 8, and Drawing EWD-47E-029, "Electrical Wiring Diagram Standby AC Power System Diesel Generator 2 Diesel Start Air Compressors DSA-C-2A1 and DSA-C-2A2."
The team reviewed the actions for this problem and found that the dispositioner had concluded that the drawing errors were an isolated eVent; however, the dispositioner had not evaluated any other drawings associated with Modification BDC 55-0729-OD.
"Upgrade EWD's for DGl & 2 Auxiliary CKT to
"Top Tier" Status,"
which revised the subject drawings.
The team noted that the drawings had been revised on January
~
1994'ut were not identified as erroneous until December 28.
1994.
The team asked how this could be an isolated case without reviewing other drawings in the design change package, especially since the design change package and the subject drawings had been reviewed; verified, and approved.
The system engineers stated that a review of all drawings verified by the person responsible for the incorrect verification of the subject drawings would be performed in order to support the conclusion that this was an isolated event.
This violation was of minor safety significance, was not reasonably expected to have been prevented by corrective actions of a previous violation, corrective actions had been initiated to correct the drawing error, and comprehensive corrective actions had been identified to prevent recurrence.
This met the criteria for discretion as detailed in 10 CFR Part 2 Appendix C
Paragraph VII.B.(2). therefore, this violation was not cited.
2.2. 1.4 Problem Evaluation Request 294-1090 This problem evaluation request was prepared to address the effects of leakage of Valve DSA-PCV-1B, a pressure control valve in the diesel starting air system, which was caused by cor rosion products.
The team found that the resolution offered and justified by the system engineer, was to crack open Valve DSA-V-723B2.
a vent downstream of the pressure control valve.
The team noted that this solution resulted in a breach of the diesel starting air pressure boundary:
however.
the system engineer's evaluation did not address the impact of this new plant configuration on the commitment for five air starts documented in the safety analysis report.
The team was informed by the engineer that this configuration existed for less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and the diesel starting air system was inspected by the operators every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
The team concluded that this was another example of a lack of rigor on the part of technical services engineers.
2.2. 1.5 Problem Evaluation Request 294-0796 This evaluation request was initiated on August 16, 1994. to document that authorized nuclear inspector review and concurrence had not been obtained for the processing of two work packages on the safety and relief valve repair check sheet.
Procedure 10.2.8.
"Testing and Repair of Safety and Relief Valves," Revision 15 'ection 6. 1, requi red the authorized nuclear inspector's review.
The dispositioner determined that the probable cause was misinterpretation of when the authorized nuclear inspector 's review was required.
For the corrective action, an interoffice memorandum was issued on September 1.
1994, which stated that the authorized nuclear inspector's review was required whenever ASNE Code Section III relief valve testing was to occur.
The team considered that these actions were adequate.
2.2. 1.6 Problem Evaluation Requests 294-0860,
-0892, and -0894 Problem Evaluation Request 294-0860, dated September, 12, 1994, was initiated to investigate and repair a failed shut check valve on the starting air system for the high pressure core spray diesel generator.
The fact that Receiver 1C could not be pressurized was discovered on August 29 '994.
The control room supervisor stated that the operability status of the diesel generator air starting system was degraded, but operable.
Part of the corrective action plan, until the check valves were replaced, was to use an operator work-around.
While pressurizing the air receivers, the operators took manual control of Compressor 1C and pressurized Receiver 1C through Receiver 2C.
Since Receiver 2C filled faster than Receiver 1C, the relief valve on Receiver ZC lifted.
The problem evaluation request stated that the problem would be corrected once the inlet restriction to Receiver 1C was removed.
Problem Evaluation Request 294-0892, dated September 29, 1994, stated that the relief valve on Receiver 2C lifted, and stuck open, causing the air receiver to depressuri ze.
The operability determination stated that the system was operable.
A licensee representative stated that the system was declared operable on the evaluation request since the request had been written after the air receiver had been repressurized.
The dispositioner determined that rust in the diesel starting air system had caused the relief valve to stick open.
In additions the system engineer determined that th'e 1724 kPa (250 psig) set pressure of the relief valve wa) too close to the normal operating pressure of the air receivers (1620 to 1689 kPa (235 to 245 psig)).-
The dispositioner stated that the opening of the relief valve was not generic to the other diesel generator air starting systems.
since the relief valve set points for the other systems were not as close to the normal operating pressures.
The evaluation request stated that there was a modification in process to upgrade the two air,receivers to a higher pressure which would allow the set points of the relief valves to be increased.
However, the team noted that the dispositioner did not address the generic impact of moisture and rust in the other starting air systems which could cause the other relief valves to stick open.
Problem Evaluation Request 294-0894, dated October 10, 1994. stated that the relief valve on Receiver 2C had lifted three times since it was rebuilt on September 29, 1994.
A modification was completed on November 15, 1994. to replace a clogged pipe leading to Receiver 1C.
This modification allowed the air receiver to be pressurized as designed, and not through Receiver 2C.
The team concluded that the impact of moisture in the starting air systems for the other diesel generator had not been considered.
The team also concluded that the operator work-around for pressurizing Receiver 1C had not been effective since it caused the additional problem of the relief valve on Receiver 2C opening and depressurizing the receiver.
The failure to evaluate the generic implications was another example of weakness in implementing procedures.
2.2. 1.7 Problem Evaluation Request 294-0506 This problem evaluation request dated September 15.
1994, was prepared to document two pressure relief valves installed backwards.
Relief Valve PSR-RV-116 was installed backward on the discharge side of post accident sampling Pump PSR-P-4.
which provided sampling capability from the suppression pool.
Relief Valve PSV-RV-118 was also installed backward on the discharge of post accident sampling Pump PSR-P-6
~ which provided sampling capability from the sump.
It was noted that in this condition the licensee concluded that the system was operable.
The team noted that the followup assessment of operability recommended that the relief valves be reoriented during the ninth refueling outage, which the plant was in at. the time.
In addition, it was recommended by the dispositioner that the work orders for the valves should be made a restart requirement.
The assessment also stated that, if the plant were placed into Modes 1,
2 or 3 without completing the work, a further assessment of operability was needed.
The valves had been inoperable since 1989, when they had been last worked on.
Relief Valve PSR-RV-116 was reoriented during the ninth refueling outage.
but Valve PSR-RV-118 was not.
On July 9, 1994, Problem Evaluation Request 294-0506 was removed from the reactor restart list with one of the relief valves installed backward.
At the time of the inspection, the relief valve was still backward.
The team determined that the additional operability assessment had not been performed prior to going into Hodes 1,
2 or 3, which had been required by Problem Evaluation Request 294-0506.
The team informed a
licensee representative of this problem and Problem Evaluation
Request 295-0163 was initiated on March 2, 1995.
The initiator of the request identified that a lack of attention to detail and failure to follow through on the corrective actions of Problem Evaluation Request 294-0506 were the probable causes; however, other licensee personnel determined that the
- corrective actions.
not scheduled for completion until September 15.
1994,
. were an internal commitment.
The team concluded that the implementation of'he corrective actions delineated in the problem evaluation report was slow.
2.2. 1.8 Problem Evaluation Request 94-1065 This problem evaluation request identified issues of relief valves not opening
.
when required and relief valves remaining open after lifting.
The licensee had 13 examples of these problems.
The initiator stated that there was no initial operability assessment since the past problem evaluation requests addressed operability.
Also, no immediate corrective actions were taken.
The requi rement not complied with was given as "possible inadequacies in the relief valve program."
The probable cause was "inadequate repair and/or trending of relief valve problems."
And, one of the recommended corrective actions was to "reevaluate the effectiveness of the relief valve program."
The team noted that the corrective actions described in the document did not address the generic concerns identified as the problem described.
Seven actions to be taken were documented; however.
these were, in general, for individual valve problems with one action to develop a relief valve data base that included water cleanliness and susceptibility to outlet corrosion.
The team found that the engineers had not realized that there were other problems with the relief valv'e program.
The engineers treated each problem as an isolated case and had not attempted to identify the root causes and correct the problems with the program.
The team, therefore, identified the concerns with the relief valve program, including those discussed in Section 2.2.3.2, as an unresolved item (397/9503-04).
2.2.1. 9 Conclusions The team found, in the resolution of problem evaluation requests.
examples of inadequate root cause determinations.
untimely corrective actions and failure to consider generic implications.
2.2.2 System Notebooks The team reviewed the system notebooks for the emergency diesel generators.
the standby service water system.
and the residual heat removal system.
While there was no regulatory requirement for system notebooks, the team reviewed these documents in an effort to determine how well the notebooks met licensee management's expectations, and the quality of effort exhibited by the system engineers.
The team was informed that the system notebooks had been in existence since January 1994.
The team found that management had determined that system notebooks could be useful documents in identifying where resources were needed as well as providing historical data on the system.
Therefore.
a memorandum was issued to di rect the creation of the notebooks.
The team was told that the notebooks were also intended to be a tool for system engineers to provide system information to new engineers when system responsibility changed.
The team found that the expectations for the content of the system notebooks had only been communicated verbally and that the system notebooks reviewed had not achieved the desi red objectives.
For example, the notebooks did not contain any information related to deficiency tags hung on system components, component failures were not included, and system performance was not trended.
Also the team found different levels of quality in the system notebooks reviewed.
The team discussed these findings with the Manage
- System Engineering.
The manager acknowledged the findings and stated that the system notebooks would be evaluated and improvements made as warranted.
The section of the system notebooks developed by the probabi listic safety analysis group was found to be excellent.
The team considered this section to be indicative of a high quality engineering product.
2.2.3 Relief Valve Program 2.2.3.1 Main Steam Safety Relief Valves The team reviewed the leakage test results of the check valves for the
.
accumulators of the automatic depressurization system main steam safety relief valves.
The purpose of the check valves was to provide a safety-related barrier between the nonsafety-related containment instrument air system and the safety-related accumulators for the seven automatic depressurization system valves.
The check valves were installed to prevent a backflow of air from the accumulators to assure that the automatic depressurization valves could be held open following fai lure of the air supply to the accumulators.
The team reviewed Procedure 7.4.0.5.53.
"CIA-V-40 and CIA-V-36 Operabil.ity Test," Revisions 4 and 5, which was written to meet the ASME Code Section XI requirements.
The team noted that Section 4.2 of the procedure stated that
"[u]se of the symbol g denotes a specific Technical Specification limit or requirement; fai lure to meet the acceptance criteria on these items requires immediate referral to the respective Technical Specification action requirement."
Also, Section 4.5 stated that
"LmJeasured test parameters beyond the action value after evaluation requires that the valve be declared inoperable."
During the test performed on June 28.
1992, four of the seven check valves had leak rates in excess of the action value provided on Attachment 9. 1,
"CIA-V-40 Series Operability Data Sheet."
The team noted that. also on Attachment 9. 1.
the leakage rate contained a note annotated with the symbol g.
The note stated that "Li]fthe calculated leakage rate was greater then
SCFH (ACTION Value)
~ refer to Precaution 4.5."
The valves that had leak rates in excess of the action value of 28.317 sLh (1 scfh) were CIA-V-40U, -40S,
-40M, and-40N.
with as-found leakage rates of 155.912 sLh (5.506 scfh).
97.948 sLh (3.459 scfh),
217. 332 sLh (7. 675 scfh). and 139. 772 sLh (4. 936 scfh).
respecti vely.
These problems were documented in Problem Evaluation Request 292-804, issued on July 1, 1992, and were designated as potentially reportable.
A reportabi lity evaluation was performed which concluded that the four failed valves were not reportable and that the automatic depressurization valves were operable.
The engineer reached this conclusion on the basis of an evaluation that indicated a leak rate of 283. 17 sLh (10 scfh)
was acceptable; however, no basis for the 283. 17 sLh was provided.
Further review of the acceptability of the basis for the higher leak rate was required.
The team identified this as an unresolved item (397/9503-05).
2.2.3.2 Pressure Relief Valve Testing and Repair The team reviewed maintenance Procedure 10.2.8, Revision 15, "Testing and Repair of Safety and Relief Valves."
The purpose of the procedure was to provide criteria for testing, repair, and modification of pressure relief valves.
The team noted that Section 4.0 of the procedure,
"Precautions and Limitations." did not have any precautions regarding determining the position of the adjusting ring(s) prior to disassembly.
The team was concerned about the lack of guidance because improper ring settings could affect valve opening and reseating characteristics.
In Section 6. 1, Valve Repair and Rework. there was a note requi ring the authorized nuclear inspector's review.
However, under Section 6.2 of the procedure.
Testing/Resetting, there was no note requiring that the authorized nuclear inspector review the testing/resetting of ASME Code Section III relief'alves.
In Section 6. 1.3.j, the disassembly section, the ring positions were to be determined by the number of turns and notches of the adjusting rings.
However, no instructions were provided in the procedure on the direction to turn the rings'r to identify a reference point for counting turns and notches.
Also, in Section 6. 1.8.d, reassembly.
instructions stated that the adjusting rings should be installed to the same position when removed or to manufacturer's specifications.
Further in the same paragraph. it was stated that the ring positions were for information only since they would be adjusted during testing.
Some of the relief valves that were installed in the plant had been flow tested by the manufacturer prior to shi pment. with the rings set for optimum performance.
including reseat pressure.
However. the mechanic was not directed to refer to the original manufacturer's test sheet for ring settings.
or to consult the manufacturer's valve instruction manual f'r the method used to measure ring position.
In a note following Section 6.2.5.b, it was stated that control'ing settings for liquid service valves had little or no influence over the valve reseat characteristics.
The team determined that this statement was incorrect.
The team reviewed the styles of liquid service valves installed in the plant and
determined that some of those valves were specifically designed for liquid service and had an adjustable ring for blowdown adjustment.
The deficiencies identified above may have been contributors to the problems identified in Problem Evaluation Request 94-1065 discussed in paragraph
. 2.2. 1.8.
Inadequate testing procedures and lack of understanding of relief valve performance could contribute to relief valves not operating as designed.
On the bases of the items identified above, the team concluded that the procedure was inadequate.
The team identified the inadequate procedure as a
violation of 10 CFR 50. Appendix B. Criterion V (397/9503-06).
The team also reviewed the in-service test results conducted for pressure relief valves in accordance with ASNE Code Section XI for a five year period from 1990 through 1994.
The team reviewed surveillance Procedure 7.4.0.5.20, Revision 7, "Testing of Technical Specification Related Safety/Relief Valves."
The purpose of this procedure was to identify the pressure relief valves that required testing in accordance with Technical Specification requirements'nd to identify the frequency and documentation requirements.
The valves were to be tested in accordance with Procedure 10.2.8
~ discussed above.
The team did not review the list of valves to determine if the licensee had included all of the applicable valves in the list.
Eighteen pressure relief valves were tested during the 1990 in-service test program.
Out of the 18 valves, 6 failed their set pressure tests.
Test summaries, which were included with the test results, contained the results of the tests and a brief description of each valve.
The team noted that four of the valves tested (RHR-RV-1A, SW-RV-1A, SW-RV-1B and RHR-RV-1B) were Crosby style JR-WR.
In the summary.
the responsible engineer stated that these valves were capable of set pressure adjustment only.
The team determined that the Crosby style JR-WR had an adjustable ring for blowdown, since the WR in the style designation stood for water ring.
In addition, the team noted that a number of valves did not have the authorized nuclear inspector's review annotated on the test sheets.
The valves without the review were:
Valves CIA-RV-5A, CIA-RV-5B, RHR-RV-30, RHR-RV-1A, RHR-RV-5, RHR-RV-SBA, and SLC-RV-29B.
The team reviewed the 1991 in-service test results for the pressure relief valves and found problems similar to those noted above for the 1990 test results.
Valve RCIC-RV-19T.
a Crosby style J0-25-WR, did not have an as-found or as-left nozzle ring position noted on the test sheet.
The mechanic marked the test sheet as "N/A." although this style of valve had an adjustable ring.
The team also found that there were no ring positions for Valves SW-RV-1A and SW-RV-lB, which were also Crosby style JR-WR.
The team noted that there was no authorized nuclear inspector review annotated on the test sheet of Valves CAC-RV-65A. CIA-RV-5A, RCIC-RV-19T, LPCS-RV-31 and SLC-RV-Z9B.
During review of the 1992 in-service test results.
the team noted that five valves.
manufactured by Crosby with an adjustable water ring, were described in the summary as only having set pressure control.
These were Valves CCH-RV-2A ~
CCH-RV-2B
~
RHR-RV-1A, RHR-RV-1B. and SW-RV-1A: all of which had adjustable rings for blowdown control.
In addition, the licensee noted in
the summary that Valves RHR-RV-1A and RHR-RV-1B were mounted horizontally by design.
The team discussed the horizontal relief valves with licensee personnel.
The responsible engineers stated that they intended to contact the valve vendor concerning the installation and evaluate necessary actions.
The team reviewed the 1993 and 1994 in-service test results and found that the test sheets for some of the valves had ring positions marked.
The team had no
'omments on the 1993 and 1994 data.
The team discussed these issues with the licensee and found that the responsible engineers did not appear to know how the valves in the in-service test program operated, or even if those valves were capable of being adjusted for blowdown.
The responsible engineers had been unaware that a horizontally mounted relief valve might not operate properly.
The team expressed a concern that the pressure relief valve adjusting ring settings might be incorrect.
Problem Evaluation Request 295-0162 was initiated prior to the end of the inspection to address this concern.
A followup assessment of operability was initiated which required an operability assessment of the ring settings for the pressure relief valves.
The team reviewed the followup assessment for operability, completed on March 8, 1995. for Problem Evaluation Request 295-0162 and determined that the engineers had only addressed the pressure relief valves in the in-service test program.
The engineers had not reviewed any safety-related air or steam service valves, or any safety-related liquid service valves that were not in the in-service test program.
The team noted that the engineers documented in the operability assessment for relief Valve RCIC-RV-19T that the maximum flow that the valve was required to pass was 125 Lpm (33 gpm) and, at the valve's set pressure of 689.5 kPa (100 psig), the rated capacity of the valve was 757 Lpm (200 gpm).
The engineers did not address that the valve might be oversized.
The team was concerned that the low flow condition of 125 Lpm (33 gpm) could be sufficiently,below the rated flow that valve chattering could occur, possibly damaging the valve.
The engineers also addressed, in the operability assessment.
Valves SLC-RV-29A and SLC-RV-29B, which were installed on the discharge side of positive displacement Pumps SLC-P-lA and SLC-P-1B for overpressure protection.
The engineers stated in the assessment that a low nozzle ring setting had a lesser consequence in terms of a reduced blowdown (i.e., the valve would reseat at a
ressure closer to the lift setpoint):
however. the assessment also contained etters from the valve manufacturers concerning ring positions.
The vendors stated in the letters that, if the ring was set too low, wher e the disc seat was too far above the port hole bottom on the ring, more overpressure would be requi red to reach rated capacity.
The engineers did not take this statement into account for Valves SLC-RV-29A and SLC-RV-29B. If the valves were required to open.
the overpressure could exceed 110K of design pressure, which would put the components being protected into an unanalyzed condition.
At the exit meetings the team was presented a revision of the followup assessment of operability.
The revision addressed the operability concerns noted above and provided additional information.
In Attachment 2 of the revised followup assessment of operability, the engineers addressed the remaining safety-related relief valves.
The engineers stated that air and gas service relief valves were not addressed in the original assessment because
"testing requirements of PPH 10.2.8 satisfactorily complete set pressure.
full flow capacity and control ring adjustments (as appropriate to achieve proper blowdown results for all gas service pressure relief devices."
The team questioned the reliance on Procedure 10.2.8 for the
.
determination of operability of relief valves since the procedure was found to be inadequate.
particularly in the setting and adjusting of the rings.
Also, in Attachment 2 of the revised assessment, the engineers provided information which indicated that Valves SLC-RV-29A and SLC-RV-298 had been replaced during the last refueling outage.
Additionally, the engineers stated that the capacity of these valves was 481 Lpm (127 gpm) with a calculated maximum demand of 155 Lpm (41 gpm).
The engineers concluded that adequate flow would be maintained.
The team noted that the engineers again did not appear to understand the relationship between starving a relief valve, resulting in valve chatter, and potential damage.
The team concluded that the relief valve program was weak and that further information was needed to verify the adequacy of the relief valve program.
Review of the maintenance records of the safety-related relief valves.
along with interviews of the maintenance personnel who performed the work, will be necessary to evaluate the operability of the safety-related relief valves.
Also, further information was needed to evaluate the acceptability of the operability assessment for Problem Evaluation Request 295-0162.
These were identified as an unresolved item (397/9503-04).
2.2.4 System Inspections The team performed inspections of the emergency diesel generators, the standby service water system, and the reactor core isolation cooling system.
The team was accompanied by the system engineer and the cognizant (design)
engineer on each inspection.
The team reviewed the applicable system notebook prior to the field inspection of the system to be able to evaluate the effectiveness of the system notebooks as well as the performance of the system engineer.
During the inspection of the emergency diesel generators, the team noted numerous deficiency tags hanging on various components of the diesel generators:
several were more than 2 years old.
The system engineer
. who has had responsibility for the emergency diesel generators for approximately 2 years, was not able to identify some components without referring to the component identification tag.
Also, the system engineer was not able to explain why the starting air tank relief valves had been changed from the original design.
other than the change had been made several years ago.
During the inspection of the standby service water system, the team noted that the system engineer.
who has had responsibility for the standby service water system for approximately 1.5 years.
was knowledgeable of the deficiency tags hung on components.
The team also noted that the system engineer had difficulty identifying some components in the system without referring to notes.
2.3 Self Assessments The team reviewed 12 reports documenting quality assurance assessments of engineering.
Host of these assessments had been performed at the request of engineering management.
The team noted that the assessors had identified many examples of fai lure to implement procedures.
Some of the procedures that were not properly implemented were Procedure NOS-9, Procedures/Instructions Control:" Procedure EI 2.8,
"Generating Facility Design Change Process" (five examples);
and, Procedure 2. 15.
"Preparation, Verification, and Approval of Calculations."
Also identified were instances where procedures were not revised and appropriate reviews were not performed during the implementation of various parts of the basic design change procedures.
The team identified an example of not performing a safety evaluation for the addition of a polisher/filter unit for the diesel fuel oil system.
A quality assurance person identified similar examples in an assessment of the safety evaluation program documented in Technical Assessment 294-001.
The technical assessment did not provide any corrective actions to ascertain the depth of this problem area.
The team concluded that this was an example of a 'tack of management attention for not ensuring the problem was properly identified and
'corrected.
One of the procedures identified in the assessments as not being properly implemented was Procedure EI 2.8.
The assessment identified a failure to implement the procedure; however. there were no corrective actions to address this failure.
The team also identified an example of failure to implement Procedure EI 2.8; also. there were no recommended corrective actions to address the failure to implement the design procedures.
The team found that while the quality assurance organization was identifying problems.
the problems were not always'ffectively corrected.
FOLLOWUP
-
ENGINEERING (92903)
3. 1 Closed Unresolved Item 397/9402-01:
Notchin Modification During a previous engineering inspection, the NRC identified a concern with the safety evaluation performed for a modification for the reactor vessel level indicating system.
In response to a request for technical assistance, the Office of Nuclear Reactor Regulation reviewed the safety evaluation for the modification of the reactor vessel level indicating system to incorporate a backfill system (referred to as the "notching modification" ) and concluded that the modification did not involve an unreviewed safety question.
On the basis of the conclusion for the technical assistance request, this item is closed.
3.2 Closed Unresolved Item 397/9402-03:
Instrument Line Orifice Diameter During a previous engineering inspection.
the NRC identified a concern with the identification of improperly sized orifices in instrument lines penetrating primary containment.
An orifice size of 0.635 cm (0.25 in) had been identified by the original design.
but orifices of up to 1.27 cm (0.5 in)
~
were found to be installed.
The team found that the disposition of this issue did not address ASNE Section III class separation requirements as provided in Regulatory Guide 1. 11,
"Instrument Lines Penetrating Primary Reactor Containment."
The team noted that the licensee had committed to comply with this regulatory guide.
However, all of the discussions in the evaluation, and the associated safety evaluation.
were related only to events analyzed in Chapter 15 of the Final Safety Analysis Report.
There was no discussion related to the applicability of Regulatory Guide 1. 11.
Although the compliance with the regulatory guide was not explicitly addressed, the licensee demonstrated that the proposed resolution was in compliance with Regulatory Guide 1. 11.
On the basis of the demonstration that. in the event of an instrument line break downstream of the orifice, releases would be within design limits and there would be sufficient make-up capability, the team concluded that there was no violation of regulatory requirements.
However, the team also concluded that the lack of evaluating the issue with respect to the regulatory guide was another example of a lack of rigor on the part of design engineering.
3.3 Closed Ins ection Followu Item 397/9402-04:
Over view of Non-Valid Problem Evaluation Re uests As documented in NRC Inspection Repor't 50-397/94-02, a concern with Plant Procedures Manual 1.3. 12,
"Problem Evaluation Report (PER)," not containing strong provisions for the overview of non-valid problem evaluation requests was identified.
Procedure 1.3. 12 contained requi rements that an individual discovering a condition adverse to quality initiate a problem evaluation request, but, if the supervisor did not agree with this assessment, the potential problem would not be entered into the problem evaluation request system.
The action of determining a problem identification as non-valid could result in an actual condition adverse to quality not being proper ly evaluated.
The team noted that this concern was addressed by revising the problem evaluation request procedure to requi re potentially non-valid problem evaluation requests to be:
forwarded to the problem evaluation request program manager:
and, retained for semi-annual quality assessment audits.
The team verified the actions.
and determined them to be adequate to address the concern.
ATTACHMENT 1 PERSONS CONTACTED AND EXIT MEETING
PERSONS CONTACTED 1. 1 Licensee Personnel
"R. Barbee, Manager, Systems Engineering
'.
Bemis, Directors'egulatory and Industry Affairs
'.
Burn, Director, Engineering
'.
Gelhaus, Manager, WNP-2 Projects
'.
Harness, Manager, Mechanical Design Engineering
'. Holder, Manager, Regulatory Special Projects
'.
Hathews, Manager, Electrical/Instrumentation and Controls Design Engineering
'.
HcDonald, Manager, Technical Services
"J. Parrish, Vice-President, Nuclear Operations
'.
Reddemann, Manager
~ Technical Services
'. Smith, Director, Quality Assurance 1 2J Swalles Plant General. Manager
"D. Swanky Manager, Licensing and Compliance
'. Williams, Nuclear Engineer, Bonneville Power Administration 1.2 NRC Personnel
'. "Beach, Director, Division of Reactor Projects
'. Callan, Regional Administrator
'. Chamberlain, Chief, Projects Branch E
2 J Clifford. Senior Project Manager
'. Collins, Acting Chief, Engineering Branch
'. Coporandy, Project Engineer
"T. Gwynn. Director. Division of Reactor Safety
'. Perkins, Director, Walnut Creek Field Office
'.
Quay, Project Director
'.
Westerman, Chief, Engineering Branch
'Attended the exit meeting conducted March 14, 1995.
'On conference call exit meeting conducted June 7,
1995.
In addition to the personnel listed above, the team contacted other personnel during this inspection period.
EXIT MEETING An exit meeting was conducted on March 14, 1995.
During this meeting, the team reviewed the scope and findings of the report.
The licensee expressed positions on some of the inspection findings presented.
Initially, a violation was identified for failure to revise Procedure 2.8.5 as part of a modification.
After additional review. the team concluded that the portion of the procedure in question was a note and the procedure was adequate as written (paragraph 2. 1. 1. 1).
The licensee also expressed disagreement with the presentation of the example of the violation for design control.
Upon further review, the team agreed that the presentation at the exit was not as clear as
it should have been, but concluded that the violation was appropriate (paragraphs 2. 1.2. 1, 2. 1.2.2.
and 2. 1.2.3).
The licensee disagreed with the identification of a violation for not reporting the automatic depressurization system check valves exceeding Technical Specifications.
The team reviewed this issue again and concluded that there was not a reporting violation; however, an unresolved item was identified for the evaluation of the acceptability of the operability evaluation (paragraph 2.2.3. 1).
Lastly, the licensee expressed disagreement with the identification of not correcting a
misoriented relief valve as a violation for failure to take corrective actions.
The team agreed with the licensee that there was no violation (paragraph 2.2.1. 7).
A subsequent exit meeting was conducted on June 7,
1995, via y teleconference, by Nr.
C. Paulk with those persons identified above.
The licensee did not express any position on the findings as presented.
The licensee did not identify, at any time as proprietary any information provided to. or reviewed by, the tea e
'P
ATTACHMENT 2 SUMMARY OF INSPECTION FINDINGS Violation 397/9503-01 was opened (Sections 1.1 and 1.2)
Violation 397/9503-02'as opened (Sections 2. 1.2. 1. 2. 1.2.2, and 2.1.2.3).
Unresolved Item 397/9503-03 was opened (Section 2. 1.2. 1).
A non-cited violation was identified (Section 2. 1.3.2).
A non-cited violation was identified (Section 2.2. 1.3).
Unresolved Item 397/9503-04 was opened (Sections 2.2. 1.8 and 2.2.3.2).
Unresolved Item 397/9503-05 was opened (Section 2.2.3. 1).
Violation 397/9503-06 was opened (Section 2.2.3.2).
Unresolved item 397/9402-01 was closed (Section 3. 1).
Unresolved Item 397/9402-03 was closed (Section 3.2).
Unresolved Item 397/9402-04 was closed (Section 3.3).