IR 05000255/1989007

Insp Rept 50-255/89-07 on 890403-0505.Violations Noted. Major Areas Inspected:Actions on Previously Identified Item, Design Control,Onsite Followup of Written Repts of Nonroutine Events & Inservice Testing of Pumps & Valves

ML18054A826
Person / Time
Site:	Palisades
Issue date:	06/23/1989
From:	Darrin Butler, Gardner R, James Gavula, Huber M, Liu W, James Smith, Westberg R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML18054A824	List: IR 05000255/1989007 ML18054A823 ML18054A825
References
50-255-89-07, 50-255-89-7, NUDOCS 8907050314
Download: ML18054A826 (47)

Text

U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-255/89007(DRS)

Docket No. 50-255 License No.. DPR-20 Licensee:

Consumers Power Company 212 West Michigan Avenue Jackson, MI 49201 Facility Name:

Palisades Nuclear Generating Plant Inspection At:

Palisades Site, Covert, Michigan and Corporate Offic~s, Jackson, Michigan Inspection Conducted:

Palisades Site April 3-7 and 17-21, 1989 Corporate Offices April 6-7 and 17-20, 1989 Region I II Offices April 24 to May 5, 1989 Inspectors: *~G-*W.~

ROlfA:'stberg, teamLeadef Winston C. Liu, Reactor Inspector David S. Butler, Reactor Inspector c P. Huber, Reactor Inspector Inspector Contractors:

John Haller, Science Applications International Company (SAIC)

8907050~: l 4 F'DR A:OOCK

JoEllen West, SAIC 6/2-2-/<g/

Date Date Date 6 /~;J./?f Date 6hz/~9 Date

• Approved By:

Ronald N. Gardner, Chief Plant Systems Section Inspection Summary:

Inspection on April 3 through May 5, 1989 (Report No. 50-255/89007.(DRS)):

Areas Inspected:

Routine announced inspection by Regional based inspectors of:

actions on previously identified items (92702); design control (37700, 37702, 37051, and 37828); onsite followup of written reports of non-routine events (92700); and inservice testing of pumps and valves (73756).

Results:

Of the four areas inspected, no violations or deviations were identified in two area Three violations were identified in the remaining areas. The first violation pertains to multiple examples of inadequate design control (Paragraphs 4.b.(2)(a) 3, (a) 4, (b) 3, (e), (f), (i), and (j); and 5.a,b,d, and e).

The second violation-pertains to examples of failure to verify the size of socket fillet welds (Paragraphs 4.b.(2)(a) 4 and 5.f).

The third violation pertains to exceeding a Technical Specification (Paragraph 5.c).

The inspection disclosed the following weaknesses:

Facility Change packages contain undocumented engineering judgements

Numerous examples of inadequate design verification

Review of ten Specification Changes and nine Facility Changes produced 19 examples of inadequate design control

Drafting errors found during review bf Facility Change Packages

Field personnel made unauthorized design changes

Use of 11 codes of convenience

The inspection noted the following strengths:

Design procedures are good

Improved performance in the pump and valve area

Electrical DBDs are good

DETAILS Persons Contacted Consumers Power Company (CPCo)

• R. D. *orosz, Engineering and Maintenance Manager

• J. G. Lewis, Technical Director

• W. E. Garrity, Engineering Manager, Energy Supply Service

• 1. K. E. Osborne, Projects Superintendent

• R. M. Brzezinski, I&C Superintendent

• D. VandeWalle, Configuration Control Manager

• 1. D. J. Malone, Licensing Analyst

• T. J. Palmisano, Systems Engineering Superintendent

• J. 0. Alderink, Staff Engineer

• K. A. Toner, Engineering Supervisor

• R. E. McCald, Planning and Administrative Director

• R. M. Hamm, Staff Engineer

• R. P. Margol, QA Administrator

• G. C. Withrow, Engineering Superintendent, Big Rock Point

• J. Pomaranski, Energy Supply Services

• W. L. Lomis, Projects Superintendent

• D. T. Perry, Staff Engineer G. W. Sleeper, Project Engineer

1. W. L. Roberts, Plant Projects

1. R. B. Jenkins, Manager, Civil and Structural Engineering

1. D. S. Riat, Project Engineer

1. Y. F. Chan, Staff Engineer Nuclear Regulatory Commission (NRC)

1. R. Cooper, Engineering Branch Chief, Division of Reactor Safety

• B. L. Burgess, Chief, Projects Section 2A

• R. N. Gardner, Chi.ef, Plant Systems Section

1. D. Danielson, Chief, Materials and Processes Section

• E. R. Swanson, Senior Resident Inspector

• J. K. Heller, Resident Inspector

• Denotes personnel who attended the exit interview on April 21, 1989 at the Palisades sit #Denotes personnel who attended the working meeting and final exit interview on May 5, 1989, at Region III office.

Actions on Previously Identified Items (92701, 92702) (Open) Open Item (255/88020-03):

The licensee advised that the two lOOA breakers of concern serving a #2AWG cable circuit (21 Amp load)

are to be replaced during the next outage anticipated to be May 198 The work orders have been issue Pending verification of the *

completed work, this item remains open.

• (Open) Open Item (255/88020-05):

This issue concerned high pres~ure air compressor motor currents. The licensee advised that the recurring compressor motor overcurrent problem has been resolved by installing a smaller diameter drive pulley, thus reducing the load on the motor and consequently motor curren The motor is now reported to run within its service factor but for a longer period of time for each compressor cycl The inspector noted that while the foregoing solves the motor overcurrent problem, it is not evident from the documentation provided that the longer running time is compatible with the maximum demand on the compressed air syste Pending verification of the above concern, this item will remain ope.

Configuration Control Project Scope This portion of the team inspection focused on the progress of the Configuration Control Project (CCP).

The Palisades Plant CCP is a multi-disciplinary project which is intended to produce a set of design documents that accurately reflects the plant's current con-figuratio The design documents will be_ used to ensure that all future modifications utilize up-to-date design informatio The project is broken into five distinct phases of wor These phases are:

(1) Engineering and Vendor Drawing Verification (2) Design Basis Reconstitution for Selected Systems (3)

System Functional Testing (4) Safety System Design Confirmation (5) Q-list Data Base Update and Validation At this point in the project, only the drawing verification and design basis reconstitution tasks have been initiate The inspector looked at only the design basis reconstitution tas Summary Consumer Power's approach to design basis reconstitution consists of consolidating supporting information for licensing bases in a single documen The resulting Design Basis Document (DBD) contains references to lengthy calculations and analyses, such as seismic considerations*,

rather than reproducing detailed calculation This document is intended to provide a single source of reference to identify critical design parameters and locate supporting information. This document will be maintained "as-configured."

Draft DBDs have been issued for the Component Cooling Water (CCW),

2400 Vac, and 480 Vac system Draft DBDs have also been issued for four non safety-related electrical system DBDs for High Pressure Safety Injection (HPSI), Low Pressure Safety Injection (LPSI), and Service Water (SW) are currently under developmen The HPSI and LPSI DBDs are being prepared by a contractor, while Consumers Power is drafting the SW DB *4 Strengths The electrical system DBDs are well-organized and provide useful information in a concise format utilizing tables and appendices to

• separate data into usable block A technical review of closed out discrepancy reports showed that conflicting or missing data is being investigated to an adequate technical level on a timely basi In fact, several deviation reports have resulted from design deficiencies identified during the reconstitutio A detailed review of component information given in the CCW DBD showed that the information accurately reflected design information (with a few exceptions), and that original su~porting design basis documents were easily retrieve Data which did not correspond to original -documents are identified in Section 3.f of this repor * Weaknesses The CCW DBD is the only mechanical -system for which a DBD has been complete The inspector reviewed a draft of the CCW DBD which had not been signed off for technical revie The document contains a great deal of information on the CCW system, but was very poorly organized and difficult to us One of the sub-tasks under the DBD reconstitution is the development of thermal-hydraulic (T/H) model A CPCo developed code, FLOWNET, was selected to model the system Models have been developed for the SW, CCW, HPSI, and LPSI system The HPSI and LPSI models are currently being benchmarke Of-concern is the fact that the T/H models are being developed prior to finalization of the DBD, and may not reflect the final DB Since the CCW DBD contains many discre-pancies which have not been resolved, the inspector was unable to assess the accuracy of the CCW mode In addition, the CCW FLOWNET model is not finalized, since sufficient data was not available for complete benchmarkin The DBD did not indicate the correct status of this mode Conclusions The inspector was very satisfied with the level of effort Consumers Power is putting into this phase of the project. Design documents are being researched to an appropriate level of detail, and discrepancy reports are routinely create However, the content and format of the Design Basis Documents varied greatly between electrical and mechanical system The inspector identified several strengths of the. safety-related electrical DBD which could be applied to the mechanical DBDs to make them more usable and less confusin At this point in the CCP, there is insufficient material to assess the adequacy of efforts to reconstitute missing design basis information for mechanical systems. Further evaluation of the Configuration Control Project at Palisades is recommended in a later phase of the design basis reconstitution task*. Use of results from the Safety System Design Confirmation task, which has not been initiated, should also be evaluate * Detailed Findings (1) Strengths of the electrical D8Ds which could be applied to the mechanical D8Ds are identified as follows:

Electrical standards are clearly identified in a separate table in the inputs for future electrical modification Mechanical CCW design standards are buried in the text and are not easily identifie *

Mechanical system descriptions and operating information are very brief and contain lengthy verbiage, which could be summarized in table *

Electrical D8Ds describe the present system description through tabular dat History and logic information is located in an appendi Mechanical CCW D8D contains system and component histories which obscure the present design basi *

Equipment data given in the appendices of electrical D8Ds identifies and supports applicable Technical Specification (TS) and FSAR informatio The CCW system flow rates given for some components do not match FSAR value Discrepancy reports were not issued for these item *

Interfacing equipment is located in a single section of the electrical system D8D The CCW system contains supporting systems in three different sections of the document in varying degrees of detai *

Electrical information is easily located through the Table of Content References to other sections of the document within the text are minima The CCW D8D contains numerous references to other sections of the document which could be eliminated if better organize *

In electrical system 080s, all text is factual and concise, or contains short references to a discrepancy repor The CCW 080 uses terms such as 11apparently 11,

11 appears to

, or 11 in the author 1s opinion 11 in reference to design points, which are vague and misleadin *

For electrical 080s, discrepancies between information given in design documents and missing design basis informa-tion is being aggressively pursue Numerous items have been closed ou When missing information cannot be located, the appropriate design basis information for current use is given to resolve the report.* The number of discrepancy reports generated by the CCW 080 reconstitution greatly exceeds the number generated by the electrical system D80, and the resolution process has just starte *

• (2)

Both electrical and mechanical DBDs should contain discrepancy report numbers in the discrepancy list to facilitate tra*cki n Discrepancies found between information given in the Component Cooling Water DBD and original design documents include the following:

The long-term shutdown heat load is given ~s 46.39 MBTU/hr on page 38 of the DB The correct value is 46.13, as given in Engineering Analysis PAL-86-083K-O The correct value appears in Table 3-1, page 10 of the DB *

Values given in Table 3-1 do not agree with values in the FSA For example, the Primary Cooling Pump Shutdown Cooling Flow is given as 360 gpm in the DBD, but is 180 gpm in the FSA The document gives some basis for the change, but a discrepancy report was not issue Shutdown cooling flow for the charging pumps is 22 gpm in the FSAR, but 32 gpm was used in the DB Some basis is given in the text on page 48, but is deemed inadequat Finally, accident flow upon SIS for the Reactor Shield Cooling HX is listed as 126 gpm and 0.20 MBTU/hr while the FSAR does not require flow during shutdow No basis is given for the chang *

No basis is given for the change in Primary Coolant Pump flow requirement from 50 gpm to 90 gpm on page 46 of the DBD.

Verification of Design (37700, 37702, 37051, and 37828) Inspection Scope This five-week special team inspection of the design control program reviewed the following areas:

verification of design assumptions, verification of design input and documentation, review of design

.calculations and methodology; compliance with codes and standards; and validity of associated 10 CFR 50.59 review In addition to document reviews and interviews, limited walkdowns and verification of as-built conditions were accomplishe The engineering areas reviewed during this inspection were mechanical, electrical, civil/structural, instrument/control, welding, thermal/

hydraulic, and computer code Detailed Inspection Findings (1) General The licensee's modification program for the Palisades plant is described by Procedures No. NODS-P08, 11 Control of Modifications, 11 No. 9.02, 11 Facility Change-Major 11, No. 9.03, 11 Facility Change-Minor,11 and No. 9.04, 11Specification Change.

11 The inspector reviewed these documents and found them acceptabl *

(2)

In order to assess the acceptability of design changes at the Palisades plant, nine facility changes and ten specification*

changes completed from October 1987 to the present were selected at random for revie Review of Facility Change (FC) Packages (a)

FC-789:

Installation of New By~ass Low Flow Rate CVs in Parallel with Existing Auxiliary Feedwater Flow Control Valves This modification added two 1 1/2 inch bypass control valves around the existing 4 inch control valves on the P-8C auxiliary feedwater pum The modification was required to provide better flow control characteristics during low flow condition The inspecto~ reviewed the following documen-tation associated with this change with regard to NRC requirements and licensee commitments:

EA-FC-789-02,_

11 Seismic Qua 1 ifi cat ion Requirements for FIC-0736A and FIC-0737A Replacement, 11 Revision 2, September 1, 198 No deficiencies or concerns were note EA-FC-789-04, 11 EI 0736 and EI 0737 Converter Support, Component Cooling Room X8400 and X8401 Conduit Supports, West Safeguards Room, 11 Revision 0, August 2, 198 The following deficiencies or concerns were noted:

For conduit support details 9, 10, 11, 12, 13, and 14, the calculations stopped at the attachment weld to the embedded stee There was no discussion regarding the adequacy of the embedded steel due to the additional loads from the-support During subsequent discussions, the analyst stated that since the loads were relatively small and the embedded steel had such a large capacity, no evaluation was require The inspector concurred with this conclusion; however, this is considered as an example of a weakness, in that it is an undocumented engineering judgemen *

Engineering Design Change (EDC) No. 789-1 eliminated conduit support detail 13 and revised the locations of conduit support details 9, 10, 11, 12, and 1 Although this change affected calculation EA-FC-789-04, the calculation was never revised to indicate that a design change had been mad During discussions with the licensee, it was indicated that the signature of the technical reviewer for the EDC was verification that the change would not invalidate the original

calculatio However, the calculation now consisted of the original analysis with an indeterminate number of EDCs that had to be included with i Since the EDC form does not list calculations as potentially affected design documents, it was uncertain whether the technical reviewer had con-sidered the effects on the original analysi In any case, there was no basis given with the EDC to justify why it would not affect the original analysi On this basis, it is considered as another example of a weakness, in that it is an undocumented engineering judgemen EA-FC-789-07, 11 Seismic Analysis of Auxiliary Feedwater Control ESSR 88714, 11 Revision 1, August 24, 198 The following discrepancies were noted in the finite element piping analysis model:

The location of new support H224 was analyzed at

11 from the 45° elbo The piping drawing (MlOl Sheet 5113) used to install the support specified a dimension of 1 1-7 1/2 11 from the elbo This difference was not noted in the calculatio *

The length of pipe between model nodes 6276 and 6282 was analyzed as 5 1 ~10 11 lon The installation drawing specifies 5 1 -6 11 lon This difference was not noted in the calculatio *

The length of pipe between model nodes 6288 and 6290 was analyzed as 1 1-11 11 lon The installation drawing specifies 2 1-2 11 lon This difference was not noted in the calculatio *

Several additional dimensional discrepancies on the new bypass piping were also not'ed between the analysis and installation drawin These dis-crepancies ranged from 1 11 to 2 1/4 11 and were considered minor by the inspecto However, none of these discrepancies were noted in the calculatio The above discrepancies are considered examples of a violation of 10 CFR 50, Appendix B, Criterion III in that the licensee failed to correctly translate the design into the drawing (255/89007-0la).

• For the south bypass loop, the Young's Modulus was specified as 27.4 E6 psi instead of 27.9 E6 ps This is equivalent to analyzing this portion of pipe with properties at 300° instead of 70°.

This discrepancy was not noted in the analysi *

The location of the center of gravity (CG) for the new bypass valves was analyzed at 19 11 from the pipe centerlin The location specified on the vendor drawing was 22

• This represents a 15% increase in the moment arm which was not noted in the calculatio In addition to the above noted discrepancies for modelling the bypass piping, other discrepancies were noted in the model of the original auxiliary feedwater pipin The inspector could not determine whether these discrepancies were inherent in the original data or whether they occurred during the transcription of the original model into the current piping analysi However, notes in the piping model stated the following:

11Bechtel analysis is a bit off from ISO here

• 11Bechtel has modeled elbows only with SIF Elbows are used here

• 11 Review ISO for pipe schedule change

These notes led the inspector to question the validity of the assumption made in the calculation concerning the correctness of the original input dat The additional discrepancies in the model of the auxiliary feedwater piping were as follows:

For flow element FE-0736, the weight of 192 lbs was modeled at node 211 instead of node 20 Although this was only a 4 1/2 11 error on a

11 pipe, the flange pair was analytically modeled with the weight concentrated at one edge instead of at the middle of the flange *

For Valve M0-0754, the 460 lb weight was modeled at the centerline of the pipe at node 26 The weight should have been specified at the valve CG at node 268, 18

out from the pipe centerlin *

The horizontal response spectra used in the analysis was inconsistent with the spectra given in Specification C-17 The spectra used was lower and not as broad as those given in the Specificatio *

Piping between the nodes 252 and 253 was modeled as 4 11, schedule 40, instead of 6

,

schedule 8 *

The above discrepancies are further examples of violation of 10 CFR 50, Appendix B, Criterion III i-n that the licensee failed to correctly translate the design into the drawing (255/89007-0lb).

The licensee committed to rerun the piping analysis in order to reconcile difference At the same time, accurate as-built dimensions were obtained from the field for inclusion in this reanalysis effor Based on the revised analysis, the licensee concluded that the installation was within the FSAR stress allowables and was acceptabl The inspector concurred with the licensee's conclusions, but had the following comments pertaining to this revised analysi First, in the revised analysis, the stress intensification factors (SIFs) were specified as 1.0 for socket welded

• fitting Based on FSAR statements, the original Code of construction was modified to incorporate the 0.75-times-the-SIF factor, which is consistent with later editions of the Cod By applying this to the 1.3 SIF

.for socket welds, the resulting SIF will be However, by specifying the SIF at 1.0 in the analysis instead of 1.3, the licensee erroneously reduced the SIF for secondary as well as primary stresse This is inconsistent with their FSAR requirements and therefore the thermal stresses *given in the reanalysis are in error by 30%.

This error will not result in an overstressed situation, but is still an erro Second, even though the FSAR specifies the inclusion of the 0.75-times-the-SIF factor, the licensee selec-tively chose to ignore the fact that the later piping codes also increased the socket welded fitting SIF from 1. 3 to Use of 11codes of convenience" is a programmatic weakness and a poor judgement relative to fundamental engineering principle Third, even though the analytical model had been revised and rechecked, the pipe schedule flange weight and Young's Modulus deficiencies previously noted were still not detected by the license In addition, Support H225 was analyzed at a location of 9 11 instead of 6.69 11 from the elbow*as shown on the as-built drawin Although these errors are minor in nature and are not considered safety significant, they are indicative of weaknesses in the licensee's design verification proces As an additional comment, the iQspector noted that the computer generated isometric plot of the system, as required by Specification M-195, was of very poor quality.

Nodal point designations and pipe routing lines were

interposed to the extent that the plot was extremely confusing and of very little valu Consumers Power Company Drawing MlOl sheet 5113, Revision 0, 11 Piping Isometric, Auxiliary Feedwater Control Valve CV-0736A and CV-0737A Bypass Piping.

The following deficiencies were noted:

The size of the fillet weld was determined by the requirements of welding specification WPS-11.21, Revision 2; however, for the socket welded fittings, the size of the fillet weld was ~ot specified on this drawin *

In reviewing the Repai~ Inspection Checklist (RIC) for the welds in question, the weld size specified is 1 1/2

• This is misleading in that this is the size of the pipe and not the size of the fillet wel In order for the welder to determine size of the fillet weld, the pipe wall thickness must be obtained and a calculation of 1.09 times the wall thickness must be performe Although this is a relatively simple calculation, it is a design function and as such must be controlle Currently, there is no documentation to demonstrate that this design activity was performed properl In addition, there are no controls in place to check and verify this design activit Failure to provide design control measures to correctly translate the fillet weld design size into the drawing is a further example of violation of 10 CFR 50, Appendix B, Criterion III (255/89007-0lc).

• A secondary aspect, associated with the socket welds, pertains to the quality control (QC)

inspection of the completed fillet weld The RIC forms have a column for 11 QC Verification 11 but for the socket welds in question, the size of the fillet welds was not inspected by Q Line

No. 16 of the RIC form, which specifies the weld, size, gap, and type of joint was marked 11 NA

(not applicable) for all the welds in question under the QC Verification colum Although all of the welds received a Nondestructive Testing (NOT) Visual Examination (VT), it is not clear if the size of the weld was verified during these examination Since the size of the socket fillet welds was not specified on the drawing, nor noted on the RIC form, the NOT 'examiner would

• have had to determine the required size in the*

same manner as previously described for the welde No notation of size nor record of the size calculation was available in the documenta-tion provided with the NOT-VT dat In addition, the VT report did not list fillet weld gauges under 11Visual Aids Used 11 giving further indication that the size of the welds was not checke As a point of clarification, it should be noted that the VT performed on the socket fillet welds was in accordance with American Welding Society (AWS) 01.1 requirement This is a structural welding code and allows portions of fillet welds to be undersized by 1/16 11 *

This is inconsistent with the requirement of ANSI 831.1, Power Piping Code which specifies minimum fillet weld size If the size of the socket fillet welds was verified by the stated VT examination, it cannot be assured that the weld meets the ANSI B3 Code requirement Failure to verify conformance of the size of the socket fillet weld with the documented welding procedure is a violation of 10 CFR 50, Appendix B, Criterion X (255/89007-02a).

• An additional aspect was associated with the size of socket fillet weld The inspector noted that the current design practice used by the licensee

is inconsistent with the original Code of construc-tio The current practice utilizes later editions of B31.l Code which specify the 1.09 times the nominal piping wall thicknes The original Code of construction required 1.25 times the nominal wall thicknes From a technical standpoint the current practice is acceptable; however, this inconsistency has not been delineated by the licensee in the FSA Pending revision of the FSAR, this item is considered open (255/89007-04).

A further concern associated with the piping installation drawing pertains to the attachment weld for a bypass piping fitting onto the existing run pip For this situation, the drawing did not specify the type of joint nor the weld reinforcement require However, the specified fitting is a 11Weldolet 11 and as such has an existing weld prep on it and requires no additional design wor Also, the size of the fillet weld cover is specified in the welding procedure for this type of full pene-tration branch line connection.. The problem arose during the review of the RIC forms for the four branch connection weld Although these are full

• penetration single bevel grocive welds, with fillet weld reinforcement, the RIC form labels these welds as 11 F.W.

11 indicating a fillet wel For Gap Thickness, the RIC form specifies 11 NA 11 which would be appropriate for a fillet weld but not for a full penetration wel Since this attachment must be a full penetration weld, there was no documentation available.to assure that the proper penetration had been achieved using the specified fillet wel Additional review by the iQspector of the NOT Examination Reports revealed another deficienc According to liquid penetrant (PT) examination report sheet No. MKV-01, welds No. 2 and No. 13 on line EBC-3-1 1/2 did not receive a PT examination as required by Technical Specification M-152(Q) "Field Fabrication and Installation of ASME Section XI Piping Modification in a Nuclear Power Plant, 11 Revision 14, September 30, 1986, Paragraph 9. Pending verification that all four branch attachment welds are full penetration welds and resolution of the PT deficiencies, this is considered an Unresolved Item (255/89007-05).

Consumers Power Company Drawing M-207, Sheet 7, 11 Piping and Instrumentation Diagram Auxiliary Feedwater System,

Revision 10, December 19, 198 This drawing was revised to incorporate the changes in the piping and alarms for FC-78 The inspector's review of the drawing disclosed that 'the pipe size for both bypass lines was erroneously indicated as 1/2 inch pipin The pipe size should have been 1 1/2 inc The pipe was verified to be 1 1/2 inch; therefore, no safety significance was attributed to this occurrenc None of the discrepancies noted above were safety significant or impacted equipment operabilit (b)

FC-722: Backup Nitrogen Supply to ESS and SWS Valve Operator Air Supply This modification added five separate nitrogen supply stations in order to provide a backup supply to 11 air operated control valve In the event of a total loss or deterioration of the normal air supply, the backup nitrogen supply will provide the necessary flow and pressure to operate the valves for the required period of tim The inspector reviewed the following documentation associated with this change with regard to NRC requirements and licensee commitments:

EA-FC 722-02, 11Sizing of N2 Distribution Lines and Cylinders, 11 December 15, 198 No adverse comments were noted during the inspectio *

EA-FC 722-03, 11Support Details for Cylinder Mountin*g.

Sheet 18 of calculation EA-FC 722-03 contained a sketch showing a typical mounting bracket for restraining the nitrogen cylinder against the wall View A-A showed a 5/8 inch diameter threaded rod attached to an angle iron with a 1/4 inch fillet wel This weld is* impossible to mak Field inspection by the inspector found that the installed weld was a flair bevel with a fillet weld cove While this is more conservative, in that additional weld metal was added, this is considered a weakness since the engineer specified an inappropriate wel In addition, the field did not reject the erroneous design information and instead implemented their own *interpretation of the informatio Although a conservative and appropriate interpretation was made, the practice of field personnel making unautho~ized design changes is not a good practice and is considered a weaknes EA-FC-722-10, 11N2 Backup Test Evaluation for Station 5,

February 27, 198 The calculation stated that the nitrogen usage rate was 32.5 psig ~P/hour based on the test results from functional test T-FC 722-501-0 However, the test results failed to account for the post test calibration shift of 5 psig for one of the pressure gauge By incorporating this additional factor, the usage rate is increased to 33.75 psig ~P/h Using the above rate in the calculation reduces the 11Actual Operating Period" from 10.3 days to 9.93 day This is below the assumed acceptance limit given in the original calculatio No safety significance was attributed to this occurrence; however, the instrument accuracy requirements specified in the test procedure were inadequate as noted belo *

Procedure No. T-FC 722-501, 11 CV Air Supply - N2

-

Backup Performance Test, 11 Revision 0, February 6, 198 Under Special Tools/Equipment, a 0-3000 psig pressure gauge is called fo The accuracy specified is +/-2% minimu This equates to a +/-60 psig accurac The acceptance criteria for three of the four nitrogen stations ranged from 24 psig to 68 psig over the four hour span of the performance tes Failure to delineate appropriate acceptance criteria is a further example of violation of 10 CFR 50, Appendix B, Criterion III Design Control (255/89007-0lq).

-Additional reviews by the inspector disclosed *

that the pressure gauges actually used had a *

specified accuracy of +/-1%.

In addition, pretest and post test calibration data indicated.that the actual accuracy was closer to +/-0.1%. _Based on this information, the performance test results were considered adequate by the inspecto *

EA-T-FC722-501-0l, 11Calculation of Acceptance Criteria for Modification Test Procedure T-FC-722-501, 11 January 13, 198 On page 2 of the calculation, it states that the total volume of gas contained in the nitrogen bottles at 2000 psig is 209 sc This value is incorrect in that it is the_usable cylinder volume as given in Calculation EA-FC 722-0 The actual volume is approximately 228 sc By using the incorrect value, the calculated acceptance critefia for pressure drops were higher and, therefore, were non conservativ Failure to provide design control measures to~

correctly translate the usable cylinder volume from the calculation to the test procedure is a further example of violation of 10 CFR 50, Appendix B, Criterion III (255/89007-0ld).

Evaluation of the above error by the inspector indicated that the effect would be a reduction in acceptable delta P by several psig. A review of the test results found that except for cylinder station 5 this would not cause any significant proble For station 5, the acceptable nitrogen usage was exceeded during the tes The test results were reviewed by the inspector during the review of Calculation EA-FC 722-1 This review, which was discussed previously in this report, concluded that the error in the calculation had no safety significanc Consumers Power Drawing M-208, Sheet 18, 11 Piping and Instrument Diagram Service Water System, 11 Revision This drawirig had been previously revised to incorporate the changes made by FC-72 For line l/2 11-JDD-16 the Backup Nitrogen Station was given as 11lA.

This is incorrect and should instead be given as 11 3B.

Although no safety significance was attributed to this drafting error, a potential exists for making an incorrect decision concerning the nitrogen backup syste None of the above noted discrepancies were safety significant or impacted equipment operabilit *

(c)

FC-789:

Auxiliary Feedwater Flow Control Modifications (Structural)

Prior to the modification, the auxiliary feedwater flow control valves were not capable of controlling flow at a rate low enough to allow continuous auxiliary feedwater flow to the steam generators during startup and hot shutdown condition This modification installed 1 1/2 11 bypass control valves around the existing 4 11 auxiliary feedwater flow control valves located in the west safeguards roo The new valves are capable of controlling low flow rates to both steam generators from motor driven auxiliary feedwater pump P-8 The inspector reviewed the modification package and the documentation related to the modification including Engineering Analysis FC-789-08, Revision 3, dated October 10, 1988, and design drawings C-271, Revision 1 and C-274, Revision The inspector identified the following concerns:

Support Nos. EB-10-H224 and EB-10-H225 were attached to the existing whip restraint The inspector noted that the design sketch specified a 3/16 11 fillet weld between the steel column and the baseplat The baseplate was identified as one inch thick based on sheet 13 and sheet 19 of engineering analysis FC-789-0 In accor-dance with American Institute of Steel Construction (AISC), Code as implemented by the licensee, the minimum size of fillet weld for the one inch baseplate shall be 5/16

• * Si nee the 3/16 11 fi 11 et we 1 ds were fie 1 d measured, such welds therefore did not meet the AISC Code require-ment Further, drawing No. C-274, Revision 5, dated January 20, 1983, was reviewe It was found that as many as 16 of these Type III whip restraints were installe *

For Type III restraints the size of the structural members was not specified other than in Item 4 of the drawing notes, which specified all wide flange shapes as W 6 x 20, unless note However, the field measure-ments revealed that W 8 structural steel shapes were installed. There was no documentation to demonstrate that the specified size could be replaced with a different siz *

For Type III, V and XV restraints the sizes of the baseplates, welds, and the anchor bolts were not specified other than a note in the Type I restrain drawing details. This note stated 11typical connection to concrete unless noted.

. The Type I restraint detail provided sizes for the welds, baseplates and the anchor bolt The weld size between the structural members an*d the basep 1 ates was 5/16 11 fi 11 e The size for the baseplates was 13 11 x 1 11 x 1 1-1 11 *

However, the

baseplates from the field measurement for the Type III restraints were 10 11 x 1 11 x 1 1-2 11, and the fillet weld from the field measurement was 3/16 11 *

There was no documentation to show that the existing baseplates. and the existing fillet welds satisfied the intent of the original desig The preceding items were discussed in detail during the May 5, 1989 working meeting and resolve The inspector had no further concern (d)

FC-731:

Reg. Guide 1. 97 Transmitters (I&C)

These transmitters are used to provide indication of plant variables that are required by the control room operating personnel during an accident situatio The inspector reviewed the pressurizer level instrument calculation (EA-FC-731-01) and the loop power supply calculation (7906-E/I-008).

The calculations were acceptabl (e)

FC-756:

HPSI Pump Miniflow Bypass Modification This modification provided greater flow capability for the High Pressure Safety Injection (HPSI) pump performance test The minimum flow recirculation piping from the discharge of both HPSI pumps has been modified by the addition of a manual bypass around the existing restriction orifice The inspector reviewed documentation associated with this mod-ification including piping stress analyses and the support evaluations for the affected system The inspector identified the following concerns:

Bechtel 1s stress isometric drawing 03378, sheet 4 of 5, Revision 1, and drawing SP-FSK-Ml93, Revision 4, showed a dimension of 29 7/8 inches between pump 66A and the elbo The as-built dimension is 13 1/2 inche Both (ADLPIPE, Inc.) ADL 1s and Bechtel 1s stress analyses used 29 7/8 inche This dimensional discrepancy was not documented during the NRC IEB 79-14 program, nor was it corrected in Bechtel 1s and ADL 1s stress analyse Further, this discrepancy is in conflict with the assumptions contained in analysis No. CS-ESSR 87-144 that purportedly demonstrated that the Bechtel drawings are correc The inspector also noted that the input data used in the modification portion of the piping system was inconsistent with as-built drawing No. 03378, sheet 4 of 5, Revision 2, as noted below:

Node Point 3100-3580 3050-3520 3110-3050 3515-3590

Input in AOL Stress_ Analysis 9.24

11.64

23.04

20.64

As-Built Drawing Dimension

11

11

11

11

The licensee reviewer was not aware of the above dimensional discrepancie Failure to correctly translate the design into the drawings is considered a further example of violation of 10 CFR 50, Criterion III (255/89007-0le).

The as-built sketch for the modification near pump 66A was sent from the site to the engineering office for revie The inspector noted that this sketch contained a dimensional erro The 2 1-6 l/2 11 dimension was incorrectly marked on the sketc This dimension was off by nine inche Failure to correctly translate the design into the drawing is considered a further example of violation of 10 CFR 50, Appendix 8, Criterion III (255/89007-0lf).

• Pipe support drawings DC1-H198.l and DC1-H196.2 contained in support calculation No. 03378 were reviewe The inspector found that one drawing showed fillet welds at the structural joints but no weld sizes were specifie The other drawing showed a 3/16 inch fillet weld with a note 11assumed.

As a result, the design bases of the welds were not adequately translated into the drawing Failure to correctly translate the design into the drawing is a further example of violation of 10 CFR 50, Appendix 8, Criterion III (255/89007-0lg).

None of the above noted discrepancies were safety significant or impacted equipment operabilit (f)

FC-731:

1.97 Transmitter Re lacement This facility change was generated to upgrade the HPSI and LPSI flow indication instrument loops to meet the Category 2 requirements of Regulatory Guide 1.9 The safety-related instruments were installed on instrument racks which were attached to the containment wall through fillet welds and bent plate The inspector reviewed documentation associated with this FC package including a final design calculation filed with the package which was identified as calculation No. 7906-CS-03, Revision 9, dated December 9, 198 The objective of the calculation was to evaluate the structural adequacy of the instruments mounted to the instrument racks and the attachment of the instrument racks to the containment wal The inspector identified the following discrepancies in the calculation:.

The analysis criteria shown on page 3 require the CG of the instruments/equipment to be considered in the seismi~

stress calculation A review of the rack support bent

plate on page 27 found that the CG of the instruments was not considered in the seismic stress calculatiori As a result, the forces and moments at the rack support attachment were inadequately calculate Failure to adequately check and verify that the analysis was performed correctly is a further example of violation of 10 CFR 50, Appendix B, Criterion III (255/89007-.0lh).

• The calculated bending stress 11 fbx 11 shown on page 27 was in erro The 5,645 psi should be 5,976 ps The checker did not identify this calculational erro *

Failure to adequately verify and check this calculation is a further example of violation of 10 CFR 50, Appendix B, Criterion III (255/89007-0li).

The torsional moment 11 Mz 11 was not included in the stress calculation, even though the moment was obvious because of eccentricity of the load acting on the supporting bent plat Further, there was no documentation to show that the existing 3/16 11 fillet welds used to attach the instrument racks to the containment wall were evaluated in accordance with the forces and moments derived from the supporting bent plate stress calculation Conse-quently, there was no assurance that the connections between the instrument racks and the containment wall were able to withstan~d the seismic load These items were discussed in detail during the May 5, 1989 working meeting and resolve The inspector had no further question *

The new loads on the rack after modification were greater than the existing loads prior to modificatio However, the existing loads were still used in the rack support seismic stress calculation No justifications were noted in the calculation This is an example of an undocumented engineering judgement and is considered an example of a program weaknes None of the above noted discrepancies were safety significant or impacted equipment operabilit (g)

FC 732:

Containment Hydrogen Monitors-Containment Isolation Valve Logic The licensee determined that the containment hydrogen monitor isolation valve logic was not single failure proo FSAR Section 6.7, "Containment Isolation System, 11 requires that control circuits which actuate automatic isolation valves arranged in a series configuration are to be completely separate, ensuring that no single failure will compromise the integrity of the containment isolation syste The licensee

modified the logic to provide containment isolation on either a left or right channel.containment isolation signa The inspector reviewed the modification wiring changes, ppst modification testing, and surveillance testin The following drawings were reviewed:

950SB9*M201, SH. 42, Sub Panels for Vertical Section Cl3L (Cl3-4), Revision 40

950SB9*M201, SH. 43, Sub Panel for Vertical Section Cl3R (Cl3-5), Revision 43

950SB9*M201, SH. 88, CllA Control Panel Section Rl Subpanels J & K Detail Wiring, Revision 10

950SB9*M201, SH. 96, CllA Control Panel Section R5 Subpanels A & B Detail Wiring, Revision 11

E-916, SH. 1, Schematic Diagram Containment Hydrogen Isolation Valves, Revision 7

E-916, SH. 2, Schematic Diagram Containment Hydrogen Isolation Valves, Revision 7 The inspector identified two drawing error Schematic diagram E-9i6, SH. 2, had the two left channel logic relay 1 s identification switched (SP-7 and SR-7) and relay 5R-6 contacts (11/12) should be (5/6).

Schematic diagram E-916, SH. 1, should be changed to include the voltage dropping resistors to the valve position light The field wiring was installed correctly as per the wiring diagram The drawing errors* did not impact testing or operabilit The licensee corrected the drawings prior to the end of the inspectio The inspector reviewed the corrections and found them to be acceptabl The inspector also reviewed the post modification and surveillance test The tests were technically accurate and provided sufficient test overlap to ensure the complete system was teste (h)

FC-567:

Core Cooling Instrumentation (I&C)

The Inadequate Core Cooling Instrumentation (ICC!) was added to comply with NUREG-0737, Item II. The inspector reviewed those portions of the modification that involved the subcooled margin monitor and the Reactor Vessel Level Monitoring System (RVLMS).

The engineering, installation, and testing of the above equipment appeared to be acceptable.

(i) I FC-567:

Core Coolin~ Instrumentation Modification (Electrical

.

NUREG-0737 and USNRC Generic Letter No. 82-28 require the installation of an instrumentation system for the detection of inadequate cooling of the reactor cor In*order to comply with this requirement, the licensee upgraded the existing subcooled margin monitor pressure transmitters, upgraded core-exit thermocouple signals and installed a reactor vessel level monitoring syste The addition of the latter has resulted in the increase of 600 VA electrical load on each of preferred AC (120V, Class lE) busses YlO and Y2 This increased loading on these two busses also increased the loading on the associated DC to AC inverters, bypass regulator and.DC system The preferred AC system, including the inverters, and the DC system are considered safety-related or Class l The inspector observed that the licensee performed calculations to analyze the impact of the increased loading on the preferred AC bus supply breakers, cabling to the preferred busses from their respective inverters and on the DC batteries; however, no calculations or analyses were evident which addressed the impact on the inverters, bypass regulator or the DC system battery charger This r.esul ted in a concern for the capability and capacity of.these Class lE systems to perform their safety-related.function The inspector concluded that the licensee had failed to employ adequate design controls during the design stage of'

the facility change in that the full impact of the increased loading was not analyze In response to the inspector 1s concern, the licensee verified the present loading on the respective inverters and battery chargers which includes the increase resulting from the instrumentation addition The two battery charger output.currents were reported to be 93 amps and 100 amp The chargers have a nameplate rating of 200 amp The inverter output currents were reported as 22 amps and 34 amp These current readings are equivalent to 2640 VA and 4080 VA at 120 volts outpu Emergency loading antic'ipated for busses YlO and Y20 as stated on Page 3 of Design Basis Document, 11 Instr. AC Sys-DBD, Rev. C-l, 11 dated December 17, 1988, is 850 VA and 1289 V The inverters are each rated at 6kVA outpu The bypass regulator is rated at 5kVA, but will be shed during a Design Basis Event (DBE) and will not be subjected to the emergency loa Thus, the licensee feels that the devices are not overloaded and will perform their intended function The inspector concurs that based on the licensee 1s reported inverter and battery charger outputs, plus the anticipated

• emergency loading, per the Design Basis document, the inverters, bypass regulator and battery chargers will not be overloade However, failure to employ adequate design controls which would have included analyses of all impacted components is a further example of violation of 10 CFR

~O, Appendix B, Criterion III (255/89007-0lj).

None of the above noted discrepancies were safety significant or impacted equipment operabilit (j) FC-760-2:

Control Room Emergency Lighting This facility change was performed to provide additional emergency lighting in the Control Roo The lighting additions were identified and evaluated as part of the licensee's Control Room Design Review performed as required by Generic Letter No. 82-33, Supplement 1 to NUREG-073 The modification added two self-contained (battery operated with a battery charger) units and two DC lighting fixtures to the existing emergency lighting systems in the Control Roo During the review of the facility change documentation, the following items of concern were identified:

Engineering analysis EA-FC-760-2-001 was performed to analyze the mounting of the lighting fixtures to be installe Section V of this document, referring to the DC lighting fixtures, states in part 11Assume the lighting fixture is rigid....

This assumption is not justified in the analysis document and, in fact, the fixture (McMasters-Carr Lampholder, Cat. No. 1700K12)

employs a swivel join The lighting fixtures are not safety-related, but mounting is considered critical since they are in the Control Room and failure could endanger personnel or safety-related device *

The engineering analysis contains a figure showing that the lighting fixture mounting has an implied critical dimension that requires verification upon installatio Evidence could not be found in the documentation that the dimension had been verifie *

Surveillance Procedure AE-5A was developed to verify operability of the self-contained emergency lighting unit The procedure addresses battery float voltage and duration of illuminatio Acceptance criteria is 7.0 + 0.1 volts and eight hours of lamp operatio Test Test-frequency is every 24 month In contrast to this, the vendor literature in the document package recommended monthly checks of the electrolyte level, specific gravity and indicator light operation.

..

In response to th~ second concern, the inspector and the licensee measured the lighting fixture critical dimension in the Control Room and determined the fixture mountings were acceptabl With regard to the third concern, the licensee advised the inspector that a request has been issued to include the vendor recommended checks in their II Peri odi.c and Predeter-mined Activity Control" (PPAC) Program with a frequency of every six month Justification for not following the vendor recommended test frequency was not give The inspector concluded that the first concern regarding the unverified assumption in EA-FC-760-2-00I is a further example of violation of IO CFR. 50, Appendix B, Criterion III (255/89007-0ik).

None of the above noted discrepancies were safety significant or impacted equipment operabilit (k)

FC-799:

Offsite Power Reliability Improvement This facility change was performed to provide power to the cooling tower busses from the unit 1 s generator outpu To accomplish this, the high voltage side of existing spare Station Power Transformer I-3 was connected directly, without a circuit breaker, to the 345 KV side of the station 1s main transforme The low v*oltage side was connected via bus supply. circuit breakers to cooling to.wer 4I60V busses IF and I Alternate feeds to these two busses remained on Startup Transformers I-I and I-3, respectivel Both fast automatic and fast manual transfer schemes have been provided for busses IF and IG between one transformer source and the othe Fast transfer is an opeh circuit or dead bus transfer without intentional time dela In addition to the above, the s~itchyard batt~ries (which were close to end-of-life), and switchyard battery chargers were relocated and replace A portion of a fire wall between the station power transformers and startup trans-formers was to be razed and a new wall erecte The fire wall work has been deleted from this facility chang Revision 7 to Chapter 8 of the plant 1s FSAR was issued in line with this facility chang This revision states:

"Station Power Transformer I-3 can be reconnected in place of Startup Transformer I-2 within three days to provide full replacement of the failed startup transformer.

During the review of the facility change documentation, the following items of concern were identified:

Station power transformer I-3 is a 22.5/25/2 - II/25/I II/25/I2/6 MVA, ~5 C/65 C unit having an impedance of

9.3% (H to X, Y) while the startup transformer 1-2 that it would replace is a 9.5/10.6 MVA 55 C/65 C uhit having an impedance of 10.84% (H to X).

Thus, if Station Power Transformer 1-3 was reconnected to the 2400V system, a higher 345KV system contribution to a 2400V system fault would be anticipate It is not evident that this impact on the 2400V system's fault withstand capability has been evaluated for this increased fault dut The increased fault duty that would be imposed on the 2400V system by reconnecting Station Power Transformer 1-3 to serve the 2400V busses requires evaluatio Also, prior to placing this transformer in this service a 10 CFR 50.59 review is require The logic for the fast automatic transfer scheme does not include a synchro-check feature, thus an out of phase transfer is possibl The inspector considers this an observation worthy of licensee review and reevaluation since induction motors and their driven loads upon out of phase transfers can be subjected to severe transient torques that may exceed design stresse ANSI Standard C50.41, Section 15, recommends that to limit the possibility of damaging the motor or driven equipment, or both, the power supply be designed so that the resultant vectorial volts per hertz between the motor residual volts per hertz and the incoming source volts_per hertz at the instant the transfer or reclosing is completed does not exceed 1.33 per unit volts per hertz on the motor rated voltage and frequency base Fast transfer between sources that are in-phase have been accepted as limiting the resultant vectorial volt~ to 1.33 per uni A review by the licensee failed to indicate that the provisions of Section 15 of ANSI C50.41 or reference to its intent were included in the procurement documentation for the cooling tower pump and fan motor Since the cooling tower pumps have shafts in excess of 20 feet and thus are likely to be more fragile than typical close-coupled motor load systems, this potentially increases the risk of shaft failure even if the voltage difference is small enough to protect the moto The licensee was unaware of any study made to evaluate or determine the magnitude of the resultant vectorial volts per hertz between motor residuals and incoming suppl The fast manual transfer scheme for transfer of cooling tower busses lF and lG is supervised by a manual synchro-check circui However, the fast automatic transfer scheme has no synchro-check feature to block transfer in the event the sources involved are out of phas *

The inspector recognizes the fact that under all planned 345KV system operating conditions the power sources involved in bus transfers at Palisades will by procedure be in-phase and thus no extreme motor/load transients should resul However, switchyard alignment is under-stood by the inspector to be under the control of the transmission system dispatcher, rather than the nuclear plant operator and the nuclear plant operator could be unaware of any phase difference The possibility exists that the two sources could be electrically separated ahd a phase difference exist such that a bus transfer damaging transient could resul None of the above noted discrepancies were safety significant or impacted equipment operabilit.

Specification Changes (SCs)

Specification change packages are used to document minor specification changes to existing plant equipmen The SC process is applied to changes to the specifications or setpoints of installed plant equipment resulting from modifications made by the equipment vendor, material substitutions and/or technical or code requirements needed to support maintenance activities or minor equipment modifications required to improve equipment/

system reliability or efficienc The SCs were reviewed to ensure that changes to the plant were accomplished according to NRC requirements, applicable codes, standards, and Consumers Power Company (CPCo) procedure The following SC packages were reviewed:

SC 86-145 SC 87-067 SC 87-069 SC 87-090 SC 87-163 SC 87-285 SC 87-344 SC 88-102 SC 88-069 Modify RGEM Controllers -

FC 2330 and FC 2346 SIRW Tank High Temperature Alarm Setpoint Change TIA 0328 and TIA 033 Replace SIRW Level Transmitter - LT 0331 Change SW Leak Detection (Containment Air Coolers) Flow Setpoint From 75 gpm to 300 gpm -

FS 088 Upgrade FW Flow Transmitters - FT 0701 and FT 070 Setpoint Change for St~rtup Ex-Core Detector HV Remova Low Temperature Over Pressure (LTOP) Setpoint Change - TS 0115 and TS 012 Replace Containment Pressure Transmitter -

PT 181 Upgrade SI Tank Pressure Transmitters - PT 0363, PT 0367, PT 0369 and PT 037 *

The following paragraphs address those SC packages that will require additional licensee action: SC 87-090 changed the Service Water (SW). l~ak detection (Technical Specification Table 4.1.3.13) setpoint from 75 gpm to 300 gp Engineering analysis No. EA-SC-87-090-1 stated that engineering judgement was the basis for the 75 gpm setpoin The 300* gpm setpoint was selected based on the total inaccuracies of the instrumentation loop times the full scale flow of the flow transmitter The EA and SC did not provide any justification to support what size of SW containment air cooler piping break could be detected by the leak detection instrument The operator response for annunciator window EK-1347, "Containment Air Coolers Service Water Leak, 11 was to close the inlet valve to each containment air cooler and check for leakag The operators may isolate the containment air coolers in response to an alarm setpoint that was not adequately verified or checked to meet the design intent of the SW leak detection syste Failure to apply design control measures for verifying or checking the adequacy of the SW leak detection setpoint change is considered a further example of violation of 10 CFR 50, Appendix B, Criterion III (255/89007-011). SC.87-163 upgraded FW flow transmitters FT-0701 and FT-0703 to Rosemount unit The supply voltage requirements for a 1151 DP transmitter is 12 Vdc to 45 Vdc (4 mA to 20 mA current loop).

The transmitter will operate within this voltage range as a function of load resistanc The load resistance for the FW flow transmitters is approximately 300 ohm The nominal supply voltage requirement for the transmitter as determined from the Rosemount functional specifications was approximately 19 Vd The licensee installed a zener diode in the seiies current loop to lower the transmitter operating voltag The inspector reviewed the SC and determined the licensee did not provide any design criteria for the zener diod~. Ih addition, the licensee did not state the power supply voltage nor did they measure the zener and transmitter operating voltages following completion of the SC:

At the request of the inspector, the licensee*measured the power supply, zener, and transmitter voltages for FW flow transmitter FT 70 In addition, the licensee also included FT 0702 (steam flow), and PT 0702 (steam flow pressure compensation) loop voltage The following voltage measurements (Vdc) were made:

Equipment N FT 0701 FT 0702 PT 0702 Transmitter 1.2 17. 62'

Zener 2.8 23.88 Power Supply 4. 5 From the above results, it appears the zeners were performing thefr intended functio The licensee indicated a total of 26 Rosemount transmitters have been installed with zeners in the series current 1 oop.

The FW flow inputs are discussed in the FSAR in Sections 7.5 and 10.2.3.3 relating to FW Regulating Systems and Section 7.2.3.2 which relates to FW flow instrumentation that provide input to the secondary plant heat balance calculatio The initial safety evaluation addressed the transmitter replacement and was revised to include the addition of.the zene However, the safety evaluation did not address the failure mechanism of the zener (shorting) and whether its failure would increase the probability of a malfunction of the FW flow loo The voltage measurements indicate that if the zener did short, the maximum voltage dropped across the transmitter would be less than 45 Vdc which should not increase the probability of a FW flow loop malfunctio However, failure to apply design control measures for verifying or checking the adequacy of the zener design is considered a further example of violation of 10 CFR 50, Appendix B, Criterion III (255/89007-0lm); and the failure to delineate appropriate acceptance cri~eria to demonstrate the zener was performing its design function is also considered a further example of violation of 10 CFR 50, Appendix B, Criterion III (255/89007-0lr). SC 87-344 changed the Low Temperature Over Pressure (LTOP) setpoints for Temperature Switch 0115 and Temperature Switch 012 The Primary Coolant System (PCS) over pressure protection system receives pressure and temperature information and acts to minimize the ~ossibility of overpressurizing the PCS at reduced temperatures by relieving through the power operated relief valves (PORVs).

The inspector discussed the LTOP system with the licensee on April 6, 198 The licensee provided the inspector with electrical schematics and the surveillance procedures which were used to adjust the pressure and temperature setpoint The surveillance procedures were reviewed in parallel by the licensee and the inspecto The licensee identified on April 10, 1989, that the LTOP pressure setpoint calibration tolerances would permit the setpoint to be left without adjustment above the Technical Specification (TS) allowable valu The procedures involved are the fo 11 owing:

M0-27A Functional Check for PCS Overpressure Protection System Setpoint 310 PSIA - Cold Shutdown/Heatup

M0-278 Functional Check of Overpres~ure Protection System Setpoint 575 PSIA - Plant Heatup

M0-27C Functional Check of PCS Overpressure Protection System Setpoint 310 PSIA - During Cooldown

M0-270 Functional Check of PCS Overpressure Protection System Setpoint 575 PSIA - Plant Operating The inspector completed the in-office review of the above procedures on April 13, 198 The inspector independently came to the same conclusions as the licensee and notified the licensee by FAX on April 13, 1989.

28 *

The inspector reviewed past performances of M0-278 and M0-27 This identified that on at least 17 occasions, the LTOP pressure setpoints were left adjusted above the TS allowable valu These are examples of violation of TS 3.1.8.1.a and TS 3.1.8.1.b (255/89007-03).

The LTOP protection is required to meet 10 CFR 50, Appendix G -

Fracture Toughness Requirements during heatup and cooldown of the reactor vesse The most nonconservative as-left setpoint permitted its associated PORV to lift 4.13 psia above the TS setpoin The smallest pressure margin (25 psia) available is at a heatup/cooldown temperature rate of 50°F (M0-27, Basis Document);

The maximum pressure instrument loop error is 14.06 psi Since 4.13 psia plus 14.96 psia is less than the pressure margin of 25 psia, the plant was being operated within its Appendix G limit SC 88-069 upgraded Safety Injection (SI) tank pressure transmitters, PT 0363, PT 0367, PT 0369, and PT 0371 to Rosemount unit This upgrade is similar to SC 87-16 The pressure channel is described in FSAR Section 6.1. The FSAR states in part, 11 the pressure of each safety injection tank is indicated in the main control room.

The analog pressure loop also provides high and low pressure alarm Redundant high and low pressure alarms are also provided by pressure switches (bistable devices).

Operations uses the pressure loop indicators (PIA 0363, PIA 0367, PIA 0369, and PIA 0371) to fulfill SI tank TS 3.3.1.b requirement that the SI tanks are pressurized to at least 200 psi The surveillance is performed according to TS Table 4.1.2 by verifying the pressure indication is between the alarm setpoint The SI tank pressure loop is further described on FSAR Figure No. 6-1 SH.l, 11 Piping and Instrument Diagram Safety Injection, Containment Spray and Shutdown Cooling System.

Even though the specific power supplies for the pressure loops were not identified on Figure No. 6-1, SH. 1, changes in the power supply output voltage could affect the operability and reliability of the pressure loo The Safety Review performed by the licensee stated SC 88-069 did not involve a change to the facility as described in the FSA The SC package did not provide any design criteria for the zener diode and did not provide the power supply output voltage that is required to correctly design for the appropriate zener voltag The following zeners were installed:

• * * *

Loop PT 0363 Loop PT 0367 Loop PT 0369 Loop PT 0371 10 Vdc 15 Vdc 15 Vdc 10 Vdc The licensee successfully calibrated each pressure loop following the zener and transmitter installatio The licensee did not verify the power supply, zener, and transmitter voltage at any time before or after declaring the SI tank pressure channels operabl The licensee obtained the voltage measurements at the request of the inspecto The following voltage measurements (Vdc) were made:

Egui12ment N Transmitter Zener Power SUJ2J2 li:

PT 0363 57.62 9.63 74.85 PT 0367 51 (calculated)

14.91 73.14 PT 0369 52.40 15.13 74.47 PT 0371 57.34 9.53 75.16 As can be seen from the above measurements, the transmitters were being operated outside their nominal operating range (14 Vdc to 45 Vdc).

The inspector discussed the operation of a Rosemount transmitter at a voltage greater than 45 Vdc with the manufacture The manufacturer indicated that the transmitter would continue to operate above 45 Vdc; however, the manufacturer did not have any data to support how long the transmitter would reliably operate above 45 Vd It appears that as the voltage at the transmitter increases, transmitter degradation will begi This effectively decreases the transmitter life and reliabilit A further concern of the inspector is the failure mode of the zener (shorted) that could go undetected and result in the transmitter having to withstand the additional zener voltage without malfunctionin The inspector reviewed the SI tank pressure loop power supply manua The Foxboro Model 610A power supply is designed to furnish power to a single electronic transmitte The nominal DC output voltage is 80 volt The manual also states that the output load resistance must be 600 ohms +10; -20 percen The SC package did not determine the load resistanc The manual provided detailed instructions to sum the input resistances of all the receivers in the loop (excluding the transmitter) and to adjust the load adjustment dial on the power supply tQ the difference between the loop resistance and 600 ohm The Rosemount 1151 GP transmitter performance specifications state that the 11power supply effect 11 is less than 0.005% per vol The inspector was concerned that in this case, a higher voltage zener will have to be used to lower the transmitter voltage (typically around 20 Vdc).

For instance,.if a 40 Vdc zener was selected and it failed (shorted), the transmitter voltage could increase to 60 Vd This could add an additional 0.2% error into the high and low setpoint calculation developed for Procedure No. RI-15A, 11Safety Injection Tank Pressure Channel Calibration.

Prior to the inspection, the licensee had no plans to monitor the zener voltages on a routine basi During the inspection, the licensee indicated they were looking into the feasibility of measuring the zener voltages on a periodic basis to ensure the zeners were performing their design function Failure to apply design control measures for verifying or checking the adequacy of the zener design is considered a further example of violation of 10 CFR 50, Appendix B, Criterion III (255/89007-0ln);.

the failure to verify and check the design by considering the affects of increased load resistance on the power supply is considered a further example of violation of 10 CFR 50, Appendix B, Criterion III (255/89007-0lo); and the failure to delineate appropriate acceptance criteria to demonstrate the zener was performing its design function

~

and acceptance criteria to properly adjust the power supply load adjustment resistor are considered further examples of violation of 10 CFR 50, Appendix B, Criterion III (255/89007-0ls).

Procedure No. 3.07, "Safety Evaluations, 11 was written to provide the gui"dance on determining the need for and proper completion of a Safety Evaluatio The SI Tank pressure 'channel is _discussed in the FSAR text and appears on FSAR Figure No. 6-1, SH. 1. - 10 C-FR 50.59, 11Changes, Tests and Experiments," requires that a Safety Evaluation be performed for changes to the facility as described in the FSA The pressure channel power supply is not explicitly described in the FSAR; however, the power supply and changes thereto will have a direct operability affect on the pressure channe The licensee answered question N Two of the Safety Review, 11 Does_ the i tern involve a change to the facility as described in the FSAR?

11, as 11 no 11 *

Consequently, a safety evaluation was not performe The addition of the zener diode created a different failure mode (shorted).

A shorted zener diode in this application will not alter the current flowing in the instrument loo The loop will.remain intact with the additional zener voltage being applied to the transmitte This configuration reduces the reliability

_of equipment identified as important to safety by operating the trans-mitter outside its normal supply voltage rang As a result, the likelihood of the transmitter to malfunction increase This also creates a different failure mode, one that could fail the transmitter as a -result of excessive supply voltag The inspector recognizes that 10 CFR 50.59 only requires safety evaluations for determinations of the existence of unreviewed safety questions for equipment described in the FSA In the above case, the description of the subject equip-ment in the FSAR was not explici Nevertheless, since the modification that installed the zener diode introduced a different equipment failure mode, a potential effect on equipment operability now exist Thus, the licensee should be sensitive to this type of problem when engaging in future modification This item has minor safety significanc Redundant pressure -switches were operable and could have alerted the operator on an abnormal SI tank high or low pressure conditio SC 88-102 upgraded containment pressure transmitter PT.1812 to a Rosemount uni The pressure loop provides indication only and is

• not required to be operable for any type of analyzed even The SC

_package did not perform a seismic evaluation for the new transmitter mounting arrangemen The transmitter is connected to a Class X penetration (Number 17) as described in FSAR Section 6.7.2. The NRC position (RG 1.29, 11Seismic Design Classification") is that systems affecting primary and secondary reactor containment whose failures during a Safe Shutdown Earthquake (SSE) could result in the release of radioactive materials should have their design requirements extended to the first seismic restraint beyond the defined boundaries (primary containment).

Those portions of structures, systems, or components that form interfaces between Seismic Category I and non-Seismic Category I features should be designed to Seismic Category I requirement In this case, the manual instrument isolation valve is always open which extends the primary containment boundary to the

transmitte Failure to apply design control measures for verifyi'ng or checking the seismic capability of the transmitter mounting is considered a further example of violation of 10 CFR 50, Appendix B, Criterion III (255/89007-0lp).

In summary, the inspector was concerned that modifications being made under the SC process have not consistently received an adequate level of engineering attention, as illustrated by the above example SC-89-72 (Deviation Report D-PAL-89-043)

This deviation report documented the undersized fillet welds on socket welded fittings for SC-89-7 This specification change was necessary to provide an interim solution to primary coolant system leakage from cold leg drain valve The change required the installation of a new length of two in~h schedule 160 pipe with a socket welded cap on each of the four loop drain Inspection of all eight socket fillet welds indicated that none of them met the Code required size of 3/8 inc During the inspector 1s review of the deviation report, there were several concerns that apparently were-not addresse First, although the corrective actions appear to recognize that the current RIC form does not give the welder sufficient information (specifically the size of the fillet weld), there was no recognition that QC did not and was not required to verify the size of the fillet wel The undersized condition was not discovered until the authorized inspector (AI)

pointed it out to the license All of the welds had been reviewed and approved by the licensee 1s program and yet the size had never been verifie This is considered another example of violation of 10 CFR 50, Appendix B, *criterion X, in that the size of the socket fillet welds was not verified (255/89007-02b).

The second concern pertains to the generic aspect of the proble The licensee appeared to recognize the programmatic weakness which contri-buted to the problem by revising the RIC form to include the specific weld siz However, there appeared to be no corrective actions directed toward reviewing previously made socket fillet welds for compliance with Code requirement Based on the added complication that the sizes of fillet welds in general apparently have not been verified under the licensee 1 s program, reviews of past work may not be necessarily limited to socket welded fittings. - Pending a review of the licensee 1s justifi-cation as to why additional inspection of previous fillet welds is not required, this is considered an Unresolved Item (255/89007-06).

Out of ten SC packages reviewed, ten examples of inadequate design control were identifie This is considered a program weaknes.

Inservice Testing (IST) of Pumps and Valves (73756)

This portion of the inspection was based on Consumers Power Company !ST program for Palisades Station including submittals dated December 28, 1988, for the pump program and April 21, 1988, for the valve program.

The inspectors found that the licensee 1s !ST program had not yet been approved by NR Therefore, the licensee 1 s programs were reviewed to determine whether the programs and relief requests were consistent with methods acceptable to NRC, and whether compliance with ASME code Section XI, Subsections !WP and !WV was achieved to the extent practica Administrative Controls Inservice testing of pumps and valves is controlled by the licensee to ensure that the appropriate testing is performed at the proper interval, that it is performed in accordance with approved procedures by qualified personnel using appropriate instruments, that the results are accurately recorded, properly analyzed, correctly stored and that trends of test results are monitored to predict and preclude failure of the tested componen (1) Administrative procedures are generated by the Inservice Inspection Section and then reviewed and approved by the Procedure Review Committee and Engineering Maintenanc (2) Technical procedures covering the performance of !ST are generated by the Inservice Inspection Section with the concurrence of Operation They are then reviewed and approved by Quality Assurance, Procedure Review Committee and Engineering Maintenanc (3) Scheduling of inservice testing of pumps and valves is performed by the Technical Specification Surveillance Program Coordinato Twice each month the coordinator interrogates the computer tabulation of equipment and test requirements and, through use of the Periodic Preplanned Activity Control System (PPACS),

generates a list of equipment which is to be tested, the tests which are to be performed, and the time at which the test must be complete Based on the information provided by the computer, the coordinator prepares the Technical Specification Surveillance Procedure, which provides the technical information from the computer and a practical translation of the time period within which the work should be performe Operations performs the work within the time 11window 11 in the schedul When a pump or valve fails to meet its !ST acceptance criteria, Operations immediately declares the component inoperabl When the test is completed, the test data are transmitted to the Inservice Inspection (!SI)

Section for analysi If any of the data is in the 11Alert

range the IS! Section initiates an order for increased surveill-ance of the componen The inspectors confirmed the increased inspection frequency imposed on equipment in the 11Alert 11 range by review of the Technical Specifications Surveillance Procedur The !ST Section records the pump and valve data and adds it to the 11 Parameter Manager 11, a computer program used in their trending and forecasting progra The licensee demonstrated a significant commitment to this syste Personnel who use it are convinced of its value.

r

'

• Training Training of Inservice Testing personnel is accomplished through the Training Coordinato The Training Coordinator maintains records of the contents of training classes provided by all training facilities, and the records of the personnel who have taken the trainin Generic skills are provided by the Muskegon Skills Center, which is a Consumers Power facilit Here the basic skills required _of all power plant employees is provided, with hands-on training provided on operating loops containing representative pumps and valve '

Plant specific skills, including work on valve operators, is provided at the South Haven Training Cente Dedicated valve maintenance personnel receive more intensive training through specialized contractors such as Chesterton (Packing), Fisher Valve (Control Valves), Farris (Relief Valves), Anchor-Darling (Valves),

and Limitorque (Valve Operators).

The use of the dedicated valve main-tenance personnel is credited with materially reducing valve maintenance rework.

Also contributing to that improvement was a more widespread training program to familiarize engineers and management with the more important facets of valve maintenanc Similar specialized training was provided for pump maintenance personnel through McNally Rotating Equipment but the effect of that training was not monitore Training records for all personnel are kept on the Empioyee Information System Computer progra In addition, each Mechanical Maintenance foreman has a copy of a Training Matrix Notebook which provides him with an outline of each employee's trainin Further guidance in assuring that properly qualified personnel are used on the job is provided by the individual work order, which contains a mandatory section on skill levels and training required to perform the wor Additional special training is provided by the foreman who completely reviews each job to be done with the.personnel involved before any work is initiated. In this way the workers are familiar with the full extent of the work to be done before they begin the jo Calibration Records of instruments used in the IST program to measure test parameters in the IST program were maintained in a data base and scheduled for calibration by Instrumentation and Control at the licensee's Jackson Headquarter Calibration of instruments such as stopwatches and vibration measuring equipment was done at the Jackson office, whereas calibration of instruments for flow or pressure gauges was done onsite in accordance with the plant's Instruments and Controls Computer Program schedul The inspector reviewed the calibration data associated with various gauges used in the performance of testing of the Service Water and Boric Acid Pump Additionally, calibration data was reviewed for charging flow instrumentation and a TK-80 vibration analyze The TK-80 vibration analyzer, ID No. 8428-00694, was calibrated on a

• y~arly frequency to a tolerance of + 5 percent, in* accordance with the Code requirement The flow and pressure gauges and the flow transmitter calibration data reviewed by the inspector were calibrated within+ 2 percent and as low as+ 0.5 percent for flow indicator FI-1347~ which is used during the-performance of the Service Water Pump tes The inspector reviewed the instrument storage and calibration controls provided for IST equipment and noted no problem These controls were adequate to ensure the required accuracy for the IST progra Pump Program Implementation The licensee 1 s pump IST program implementation was inspected to verify compliance with Appendix B of 10 CFR 50; 10 CFR Part 50.55a(g); and subsection IWP of Section XI of the ASME Code (1983 Edition with Addenda through Summer 1983). The inspection included a review of administrative controls, selected surveillance procedures, test results, and documentatio (1)

Program/Relief Requests The inspector reviewed the licensee 1s controlling procedure governing the conduct of IST, including associated relief request Due to the fact that approval of the licensee 1s program had not yet been granted, the inspector evaluated the program and requests to determine the extent to which compliance with code requirements was achieve To the extent practical, the licensee was meeting the code requirement However, some concerns were noted and are detailed below:

Two relief requests in the licensee 1s program were intended to provide relief from the requirements of Table IWP-3100-2, 11 Allowable Ranges of Test Quantities, 11 when the instrumenta-tion used by the licensee to perform IST, although calibrated within the requirements of Table IWP-4110-1, 11Acceptable Instrument Accuracy, 11 allowed the test results to fall outside of an allowable range, into either the 11Alert 11 or 11 Required Action 11 rang This was a nonconservative approach to be used by the licensee in the event that increased testing, due to test failures, was required due to Code allowable inaccuracies of the measuring and test equipmen This was not an acceptable practice and the inspector discussed this with the license Blanket relief was not the intent of the licensee and instrumentation used for IST purposes is within the accuracy limits specified in Table IWP-4110- Modifications to systems with inaccurate gauges were initiated by the licensee, and the licensee stated that these relief requests would be withdraw *

The method of vibration analysis used by the licensee was displacemen The inspector noted that other techniques

• such as vibration analysis using velocity measurements allows a more comprehensive analysis of the pump conditio The licensee stated that they were reviewing this techniqu During the conduct of this inspection, -NRC fssued Generic Letter (GL) No. 89-04, 11Guidance on Developing Acceptable Inservice Testing Programs 11, dated April 3, 198 Many of the. issues

.

noted above will be addressed by the licensee in their response to the G It is the licensee 1 s intent to delete the majority of relief requests contained in the ~rogram, which will require the modifications as mentioned abov (2) Completed Surveillance Review The inspector reviewed several procedures to ensure that Code requirements were met and to evaluate the effectiveness of the progra The following surveillance packages were reviewed:

Procedure No. M0-38, Revision 2, dated November 7, 1988, 11Auxiliary Feedwater System Pumps, Inservice Test Procedure, 11 performed on March 13, 198 *

Procedure No~ Q0-19, Revision 3, dated October 20, 1988, 11 Inservice Test Procedure - High Pressure Safety Injection Pumps and ESS Check Valve Operability Test, 11 performed March 8, 198 *

Procedure No. Q0-20, Revision 2, dated July 6, 1988, 11 Inservice Test Procedure - Low Pressure Safety Injection Pumps, 11 performed January 11, 198 *

Procedure No. Q0-18, Revision 3, dated October 7, 1988, 11 Inservice Test Procedure - Concentrated Boric Acid Pumps,

performed October 29, 1988, November 4, 1988, and March 8, 198 '

Special Test Procedure T-235, Revision 0, dated March 14, 1987, 11Concentrated Boric Acid Pumps **P-56A and **P-56B Performance, 11 performed March 14, 198 One administrative problem was noted in Procedure No. M0-3 Step 3.6.1.b-allowed for a vibration instrument accuracy of! 10 percent, which is outside of the allowable accuracy range specified in the Cod The instrument used during the surveillance was calibrated within the Code specified accuracy range and all of these type of vibration monitoring instruments were calibrated to

+ 5 percent accuracy as allowed by the Cod A procedure change

~as issued by the licensee on April 7, 1989, to revise the allowable accuracy from+ 10 percent to+ 5 percent accurac No other problems were n~ted.

• The licensee recently completed a modification on the HPSI system, which allowed tests to be run at substantial flow condition No problems were noted with Procedure No. Q0-1 The LPSI system pumps, tested in surveillance Procedure No. Q0-20, currently do not have a configuration that allows for tests to be conducted at or near design flow condition However, the licensee has developed preliminary plans to address this proble These plans are to install appropriate flow and pressure gauges in current system piping or modify the configuration to add a recirculation loop that would allow for substantial flow testin This action will also be addressed by the li~ensee as part of the response to GL 89-0 The inspector noted one discrepancy during the review of the surveillances performed using Procedure No. Q0-1 As part of an IST pump inspection,.reference values are to be established to compare the measured values obtained during subsequent tests to allow for comparison in order to determine the pump hydraulic conditio The values are to be measured after either the reference flow rate or differential pressure is established, as required by the Cod The licensee does not have the instrumen-tation in the line used to test the Boric Acid pumps to measure flow rate or differential pressure, and therefore could not establish the appropriate reference for testin A relief request was submitted to the NRC; however, the inspector noted that this was unacceptabl The licensee tested these pumps at design flows and pressures in 1987, and the pumps performed acceptably. In addition, the licensee noted that the reference values need to be established to fully evaluate the hydraulic condition of the pum It is the intent of the planned modifications to install the appropriate means to conduct this type of testin The inspector noted that this condition existed only for the Boric Acid and Component Cooling Water pumps (for which similar actions are being taken) and is not a concern for other pump (3) Test Observation The inspector witnessed the performance of inservice testing of the Service Water Pump The licensee uses Operations personnel to perform all aspects of the testing, including the pump vibration measuremen Vibration data was obtained using calibrated equipmen The points used for measurement were clearly marked on the pum Reference flow was estab-1 ished in the recently installed bypass header, installed to facilitate the pump testin However, the flow gauge used, FI-1347, was difficult to read, in that it swung approximately

~ 300 gpm from the desired average flo A deviation report was issued when the licensee discovered this condition during a previous surveillance and initiated

action to have the.situation corrected by June 198 The work was done in a professional manner and the Operations staff was knowledgeabl No other problems were note Valve Program Implementation The licensee's-valve !ST program implementation was inspected to verify compliance with Appendix B of 10 CFR 50, 10 CFR Part 50.55a(g);

and subsection !WV of Section XI of the ASME Code (1983 Edition with Addenda through Summer 1983). The inspection included a review of administrative controls, selected surveillance procedures, test results, and doc~mentatio (1) Program/Relief Requests As previously indicated, approval of the licensee's program had not yet been granted, so the inspectors evaluated the program and the related relief requests with respect to the guidance available in ASME Section XI and Generic Letter 89-0 Several anomalies were observe (a) Relief Request No. 2 proposed, as an alternative to full flow testing of check vilves, the partial stroke exercise during hot shutdowns and disassembly and verification of freedom of disk motion on a five year basi That is, two valves would be inspected every five years and all four would be inspected in each ten year interva This request conflicts with the NRC position on "Alternative to Full Flow Testing of Check Valves" indicated in Attachment 1 to Generic Lett~r 89-04, which states "Extension

~f the valve disassembly/inspection interval from that allowed by the Code (Quarterly or cold shutdown frequency)

to longer than once every six years is a substantial change which may not be justified by the valve failure rate data for all valve groupings."

The attachment lists three pre-requisites for reducing inspection frequency based on valve inspection experienc *

The license~ indicated that the alternative proposal in the relief request wo~ld be abandoned and that testing would be modified to reflect the intent of Generic Letter 89-0 (b)

The inservice testing of plant valves, Procedure EM-09-02 Revision 12, dated April 21, 1988 (The IST Program),

included two valve lists: Attachment 1: Valves tested by P&ID and Attachment 2:

Valve Reference List in alph~

numeric orde These valve lists contained erroneous data:

References to Relief Request (RR)-11 should be to RR-10

References to RR-13 should be to RR-12

References to Drawing Coordinates are incorrec For example:

CV-0884 on M208 lA should be G-3 instead of D-6 CV-0885 on M208 lA should be F-3 instead of D-5 Trending records reflect similar anomalous references, which are probably the result of revisions in the number of relief requests and in the redrawing of some P&ID (c) The program for IST states, in paragraph 5.2.4.C, 11Valve leakrate testing other than containment isolation valves shall be performed in accordance with IWV-342 EM-09-02 Revision 12 has no valves meeting this requirement.

However, NRC 1s 110rder For Modification of License Concerning Primary Coolant System Pressure Isolation Valves 11 dated April 2, 1981, included revised Technical Specification pages 4-17 and 4-1 Page 4-17 included the following information:

Technical Specification paragraph number 4.3.h.:

Periodic leakage testing (a), (b) on each check valve listed in Table 4.3.l shall be accomplished prior to returning to the Power Operation Condition after every time the plant has been placed in the Refueling Shutdown Condition, or the Cold Shutdown Condition for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if such testing has not been accomplished within the previous 9 months, and prior to returning the check valves to service after maintenance, repair or replacement work is performed on the valve The licensee has confirmed that all necessary tests were performed on Pressure Isolation Valves even though the tests were not included in the Inservice Testing Progra Test results for the leak testing of these valves after each refueling shutdown and cold shutdown were located and were reviewed by the inspecto The Licensee indicated that the current procedures already reflect the intent of Position 8 of Generic Letter (GL) 89-0 A sample of surveillance procedures was reviewed to confirm this statement. Several of the procedures did not provide clear guidance in this area. Paragraph 6.2 of procedure Q0-05 Revision 30, for example, indicates that corrective action shall be as specified in Palisades Administrative Procedure No. 9.23. The corrective action in that document provides a definition of LCO initiation time that does not meet the intent of Position 8. It also includes guidance for continued use of equipment not meeting test acceptance criteria that does not conform to the intent of Position 8. Similarly, Paragraph 6.0 of Procedure Q0-21, Revision 4, does not require that pumps or valves which fail to meet acceptance

4.e.(3)

criteria be declared inoperabl Stroke Time Reduction Resulting From Limit Switch Shift Experience with previous licensee programs disclosed a problem which occurs when limit switches are repositioned for optimum torque switch bypas When the limit switch controlling torque switch-bypass also controls the position indicating light, there is a potential for losing control of stroke timin Ideally, the limit switches for position indicating lights should be adjusted to operate near to the open and closed position The limit switch for the torque switch bypass is commonly set as high as 20 percent off the closed positio When the same rotor and the valve operator is used to operate both switches, there is a conflict in objective The problem can be resolved on a four rotor limit switch operator by shifting the lights to the other rotor on the closed end of trave However, on a two rotor operator, there is no simple resolution to the proble When the rotor is fixed at 20 percent off the closed position an error is automatically introduced into the normal stroke timing procedur Ordinarily, stroke timing is performed from the Control Room and the timing covers the interval between the initiation of the switch (in this case, the 11 close 11 switch) and the operation of the light at the 11 close 11 end of the stroke. If the light operates 20 percent before the end of the stroke, the timing stops 20 percent before the end of the strok Thus a stroke time that violates the 11 required action 11 acceptance standard by up to 20 percent would be acceptable in this syste As a consequence, the stroke timing would fail to comply with the requirements of IWV-3413(b), in that stroke timing would not be measured within 10 seconds or 10 percent of the specified limiting stroke tim The licensee recognized the potentially detrimental aspects of this condition when adjusting switches in response to !EB 85-0 Two rotor MOV's which required additional switches to permit separate control of position indication lights and torque bypass switches were replaced by four rotor MOV's to facilitate this change and prevent the conflicting requirement The licensee indicated that all limit switches for position indication lights are now located near the extremes of stroke travel and provide an accurate and effective means for measuring stroke tim Discussion of the !ST Program The anomalies cited in the !ST program and its relief requests are not identified as violations because the program and relief requests were not previously approved by the NR Had these documents been reviewed, these anomalies would have been identified and corrected before the documents were returne In effect, the current review has performed a similar (although more superficial) functio It provides minimal guidance to the licensee in revising the program to conform to the guidelines of Generic Letter 89-0.

The licensee has demonstrated a commendable attitude toward improving performance in the areas of pumps -and valve The training previously described represents only one facet of their approach to this are Another is the Valve Improvement Program, which sought to improve valve performance through improvements in packing selection and application, project coordination, tool application, and training of dedicated group That program resulted in the following* results:

Reduced valve maintenance rework from over 10 percent to under 1 percen *

Tripled the number of valves that could be repaired in one yea *

Essentially eliminated packing leaks on repacked valve The success of the program is attributed to the application of innovative approaches and to the cooperation of all levels of management to achieve the desired end result Onsite Followup of Written Reports of Non-Routine Events (92700)

(Closed) LER 255/88021:

Potential for the loss of the Service Water (SW)

pump Since February 5, 1987, the plant has experienced several unexplained service water pump trip The licensee determined in November 1988 that the cause of the spurious tripping was the result of the high dropout (HOO) overcurrent relay not resetting during high SW load condition The load increases were initiating the time over current (TOC) relay and along with the unreset HOO relay would trip the pump.

The licensee backfiled the SW pump impellers in late 198 This increased the SW pump capacity sufficiently to increase and maintain the motor running current above the HOO relay reset poin The backfiling increased the pump horsepower requirements from 350 Hp to approximately 375 H The motor is rated at 350 Hp and has a service factor of 1.1 The motor may be reliably operated to 402.5 Hp (assuming no losses).

Motor insulation systems are susceptible to heat buildu The SW motor has a Class B insulation syste According to ANSI Standard C50.41-1982,

American National Standard For Polyphase Induction Motors for Power Generating Stations, 11 the temperature rise of a Class B insulation system is acceptable provided the temperature-rise does not exceed 90°C as determined by the resistance method of temperature determinatio The licensee determined the temperature-rise (resistance method) was 85.86°C and at a motor efficiency of 92% would produce 375.07 H As a result of the above, the licensee considers the motors qualified for their intended us The inspector reviewed internal CPCo correspondence KAS 01-87, dated January 7, 198 The correspondence stated that 11The Plant should be advised to proceed with their plans to replace the pumps and motors with those of greater capacity as they intende While this is not necessitated by current conditions, it would be prudent for the long term.

ANSI Standard C50.41-1982 in Section 9.3.2, 11Temperature-Rise, 11 supports the above statemen The Standard states that, "Operation at the temperature-ri se values given in Table 2 for a 1.15 service-factor load causes the motor insulation to age thermally at approximately twice the rate that

t'

• y occurs at the temperature-rise values given in Table 1 for a motor with. a 1.0 service-factor load; that is, operating one hour at specified 1.15 service-factor temperature-rise values is approximately equivalent to operating two hours at the temperature-rise values specified for a motor with a 1.0 service-factor.

The inspector reviewed the operator's response to Annunciator*Number 37, 11Service Water Pump P-78 Overload/Trip.

The response was "Check relays if pump trippe If pump did not trip, then overload relay caused alarm in this case, monitor motor current and if possible, reduce service water load Pump will trip if current reaches 114 to 126 amps.

Onshift personnel indicated they would not operate the pump with the alarm presen They would immediately start the standby pump or equalize the flow between the running pumps to reduce the motor curren The inspector reviewed the SW pump's operating history for the summer of 198 The following current readings (Amps) were obtained from the 'B' shift (day shift):

SW Pump SW Pump Date A B c Date A B c 6/1 x 82 83 7/21 x 81 83 6/7 x 81 82 8/1 77 78 79 6/14 79 83 x 8/4 86 87 86 6/21 83 86 x 8/8 87 88 79 6/28 80 80 83 8/9 81 82 75 7/7 x 83 84 8/15 81 83 76 7/14 x 86 86 X Denotes pump not running None of the pumps was operated near their 1.15 service-factor current of 96 Amp The inspector advised the licensee to continue to closely monitor the SW motor currents and take appropriate measures to ensure the motors are being operated at less than 96 Amp The inspector had no further concerns on this item at this tim.

Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involves some action on the part of the NRC or licensee or bot An open item disclosed during this inspection is discussed in Paragraph 4.b.(2)(a) ~* Unresolved Items An unresolved item is a matter about which more information is required in order to ascertain whether it is an acceptable item, an open item~ a deviation, or a violatio Unresolved items disclosed during this inspection are discussed in Paragraphs 4.b.(2)(a) ~and *

1 Exit Meetings The inspectors met with licensee representatives (denoted in Paragraph 1)

on April 21 and May 5, 1989 to discuss the scope and findings of the inspectio In addition, the inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspectio The licensee did not identify any such documents/processes as proprietary.

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CPCo R. M. Brzezinski R. J. Corbett D. D. Crabtree G. J. Daggett T. c. Duffy M. A. Ferens E. Feury R. M. Hamm L. H. Keller D. M. Kennedy J. A. Mei ncke M. T. Nordin M. D. Paschke D. T. Perry u. R. Peterson w. L. Roberts R. L. Scudder G. w. Sleeper T. J. Swi eci cki D. Vandewalle R. s. Westerhof CPCo Jackson G. J. Brock Y. F. Chan R. T. DesJardins G. w. Foster B. L. Ha rs he P. Papaioannou R. Pienkos D. J. Radzwion R. c. Schmid K. A. Stevens J. L. Topper M. R. Wade G. A. Washburn K. Yeaber Bechtel M. Mau ATTACHMENT A:

PERSONNEL CONTACTED Instrument and Control (I&C)

Project Engineer System Engineer System Engineer Reactor Engineer I&C Training Project Engineer Staff Engineer I&C Reactor Engineer Supervisor, Electrical Systems Project Engineer Staff Engineer I&C Project Engineer Training Project Engineer I&C CCP Manager I&C Senior Engineer Staff Engineer Staff Engineer Senior Engineer Staff Engineer Staff Engineer Senior Engineer

. Senior Engineer Engineer (Contract)

Staff Engineer Staff Engineer Section Head, Projects Engineering and Construction Engineer (Contract)

Staff Engineer Engineer

Modification FC-567 FC-722 FC-731 FC-732 FC-756 FC-760-02 FC-789 FC-799 FC-811

ATTACHMENT B:

FACILITY CHANGE REVIEW

~

Major Minor Major Minor Minor Minor Minor Major Minor Description Core Cooling Instrumentation Nitrogen Backup Supply to Sever~l Valves Reg. Guide 1.97 Transmitters Containment Hydrogen Monitors Containment Isolation Valve Logic HPSI Pump Recirculation Path Miniflow Orifice Bypass Valves Control Room Emergency Lighting Installation of New Bypass Low Flow Rate CV 1s in Parallel With Existing AFW Control Valves Offsite Power Phase I - Repowering of Cooling Towers via the Installation/Hookup of Station Power Transformer 1-3 Installation of SW Pump Instrumentation Phase I

Procedure N NODS-P08 3.07 9.01 9.02 9.03 9.04 9.05 9.11 9.30 13.01 8303-501 8303-502 AE-5 AE-5A GOP 2 GOP 9 M0-27A M0-278 M0-27C M0-27D RI-15A RI-158 RI-18 RI-27 RI-38 RI-59 R0-11 ATTACHMENT C:

PROCEDURE REVIEW Title Revision Control of Modifications

Safety Evaluation

• * 2 Request for Plant Modification

Facility Change - Major

Facility Change - Minor

Specification Changes

Modification Procedures and Construction

Work Packages Engineering Analysis

Q-List

Identification and Tracking of CCP Discrepancies

Func_ti on Check Test-345 KV

Switchyard Battery Chargers Preoperational Test - 4160V

Susses lF and lG Breakers Basis Document for DC Lighting Test -

Turbine, Auxiliary, Feedwater Purity and Service Buildings Basis Document for Emergency Lighting Unit Duration Test and Circuit Adjustments O

Plant Heatup (Cold Shutdown to Hot Shutdown); Step 2.29

Plant Cooldown From Hot Standby/

Shutdown; Step 2.10

Function Check of PCS Overpressure Protection System Setpoint 310 PSIA-Cold Shutdown/Heatup

Function Check of PCS Overpressure Protection Setpoint 575 PSIA-Plant Heatup

Function Check of PCS Overpressure Protection System Setpoint 310 PSIA-During Cooldown

Function Check of PCS Overpressure Protection System Setpoint 575 PSIA-Plant Operating

Safety Injection Tank Pressure Channel Calibration

Safety Injection Tank Pressure Switch Calibration

SIRW Tank Temperature Indicator Calibration Procedure

Containment Service Water Break Detection

SIRW Tank Level Instrument Calibration

Calibration of PCS Overpressure Protection System

Containment High Radiation Test

• .>

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Attachment C

R0-12 Containment High Pressure (CHP)

Spray System Tests T-FC-732-501 Test Procedure for Modification to Hydrogen Monitor Containment Isolation Project Plan for Configuration Control Project

0 2