ML19249C071

Responds to Forwarding Const Worker Affidavits Re Containment Isolation Valves.Discusses Encl Insp Repts Generated as Result of Review of Affidavits.Final Rept Will Be Provided

ML19249C071
Person / Time
Site:	Zimmer
Issue date:	07/31/1979
From:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	Rosenblum D OHIO, STATE OF
References
NUDOCS 7909070275
Download: ML19249C071 (2)

Text

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NUCLEAR REGULATORY COMMISSION REGION lil g

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%.'.b/ [g GLEN ELLYN, ILLINOls 60137 JUL 3 1 1979 Mr. Donn D. Rosenblum State of Ohio Public Utilities Commission 180 East Broad St reet Columbus, Ohio 43215

Dear Mr. Rosenblum:

Thank you for your letter of June 20, 1979, transmitting affidavits f rom construction workers at the Zimmer Nuclear Power plant site. We had pre-viously received these af fidavits and the Region III staf f has been looking into the concerns expressed by the workers who made the affidavits.

In addition, these issues have been transmitted to the Hearing Board for the Zimmer plant and have been, or will be discussed during the hearing process. Please find attached the inspection reports generated following review of these matters and statements made to the Hearing Board.

At this time, the comments regarding metal particles have not been fully resolved, and our review is still in progress. We will send you a copy of the report when our review is completed.

We responded to Mr. Martin following our review of the concerns expressed in his affidavit, and explained our findings to him.

Mr. Martin's comment regardino containment isolation apparently refers to the findings in Inspection and Enforcement Inspection Report No. 50-358/79-09 (Section 10, which includes information regarding containment isolation valves at the Zimmer plant. The licensed design for the plant does allow the use of a single isolation valve for some piping, as detailed in the inspection report.

We believe that our review of each issue was of sufficient depth to identify any safety problem. However, we will respond to any further concerns brought to our attention to assure that we have not overlooked any signifi-cant deficiencies.

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3 1 I373 Donn D. Rosenblum 2

If you have further questions regarding these matters, or regarding the attached reports, please feel free to contact our office.

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Sincerely,

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o.By James G. Keppler Director

Enclosures:

1.

Inspection Rpt. No. 358/79-06 2.

Inspect ion Rpt. Pb. 358/79-09 3.

Statements made by T. Vandel 4.

Statements made by F. Maura cc w/ enclosures:

Thomas D. Martin Robert S. Ryan, Ohio Department of Energy J. R. Schott, Plant Superintendent Central Files Reproduction Unit NRC 20b PDR Local PDR NSIC TIC Harold W. Kohn, Ohio Power Siting Commission Citizens Against a Radioactive Environment Helen W. Evans, State of Ohio C. ) !,

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P' MAY 161979 M

Docket No.50-35S 'd

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Cincinnati Cas and Electric Conpany ATTN:

Mr. Earl A. Borgnann Vice President Engineering Services and Electrical Production 139 tast 4th Street Cincinnati, O!!

45201 Centleuen:

This refers to the inspection conducted by Messrs. F. A. Maura, R. U. Dettenmeier and B. M. K. Wong of this office on February 27-28,

!! arch 1-2, 19-23 and April 9-11, 1979, of activities at the L'c:. H. Zimoer Nuclear Power Station authorized by NRC Construction Peruit No. CPPR-P8 and to the discussion of our findings with Messrs.

Schott and Schwiers on March 23 and Messrs. Culvert and Schwiers on April 11, 1979, at the conclusion of the inspection.

The enclosed copy of our inspection report identifies areas examined during the inspection. Within these areas, the inspection consisted of a selective examination of procedures and representative records, observations, and interviews with personnel.

During this inspection, certain of your activities appeared to be in noncompliance with NRC requirements, as described in the enclosed Appendix A.

Regarding items 1.c, and 3, the inspection showed that action was taken to correct the identified noncompliance and to prevent recurrence. Consequently, no reply to these items of non-compliance is required and we have no further questions regarding L

these items at this time.

9 E-v' t.

Recognizing that the Zimmer preoperational test proBran is in the

%H beginning stages, we are becor:ing increasingly concerned with certain

>_c:jM cd elements of your test program controls and the discipline of opera-i i c--

tions associated with the testing. The thrust of our concern is the 8

continuing finding of munerous examples of procedure violations and inadequate comunmications, some of which have resulte.' in equipment L,

deEradation and damage. We believe prompt management action is needed to strengthen the overview and discipline of your preopera-tional test program. In this regard, please include in your response to this letter what actions you plan to take to assure test procedures are followed. Your response and our subsequent inspection of these OFF8ct >

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NRC Form alSA (R2II) (5-76) NRCM 02040

.U. S. GOVEJtyENT PRINTING oFF8CE:

1978-253-8' c -) ?

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y Cincinnati Cas and 2-Electric Company MAY 161979 P

activities will be considered in determining whether additional enforcement action is needed.

This notice is sent to you pursuant to the provisions of S etion 2.201 of the NRC's " Rules of Practice," Part 2, Title 10 Code of Federal Regulations.

Section 2.201 requires you to submit to this office within twenty days of your receipt of this notice a written statenent or explanation in reply, including for each item of noncenpliance:

(1) corrective action taken and the results achieved; (2) corrective action to be taken to avoid further noncompliance; and (3) the date when full compliance will be achieved.

In accordance with Section 2.790 of the NRC's " Rules of Practice,"

Part 2, Title 10 Code of Federal Regulations, a copy of this letter, the enclosures, and your response to this letter will be placed in the NRC's Public Document Room, except as follows.

If the enclosures contain information that you or your contractors believe to be proprietary, you must apply in writing to this office, within twenty days of your receipt of this letter, to withhold such information fron public disclosure. The application must include a full statement of the reasons for which the information is considered proprietary, and should be prepared so that proprietary information identified in the application is contained in an enclosure to the application.

Uc will gladly discuss any questions you have concerning this inspection.

Sincerely, James C. Keppler Director

Enclosures:

1.

Appendix A. Notice of Violation 2.

IE Inspection Ppt No. 50-358/79-06 cc w/encls:

J. R. Schott, Plant Superintendent Central Files Reproduction Unit NRC 20b PDR Local PDR NSIC t

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APPENDIX A ROTICE OF VIOLATION Cincinnati Cas and Docket No. 50-358 Electric Company

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Based on the inspection conducted February 27-28, March 1, 2, 19-23 and April 9-11, 1979, it appears that certain of your activities were in noncompliance with the NRC requirements as noted below. Items 1 and 2 are infractions and item 3 is a deficiency.

1.

10 CFR Part 50, Appendix B, Criterion V, states in part that,

" Activities affecting quality....shall be accomplished in accordance with instructions, procedures or drawings."

Paragraph 17.1.5.1 of the FSAR states in part that, "Acti-vities affecting quality of the facility are accomplished in accordance with written instructions, procedures or drawings...."

Paragraph 5.2 of Startup Manual Procedure SU.ACP.16 a.

states in part that all safety related equipment shall be maintained free of excessive oil, water or other material which could prevent the equipment from per-forming its intended safety functions.

Contrary to the requirements of SU.ACP.16 above, the licensee did not maintain the immediate enclosed area around the IB and Spare 125VDC Battery Chargers clean of debris and trash, and did not maintain the internals of the Epare 125VDC Battery Charger free of insulation and concrete pieces leaving a side panel off even though it had been turned over for preoperational testing.

Contrary to the requirements of SU.ACP.16 above, on March 17, 1979, the licensee did not waintain control (cg of flushing waters which sprayed down the HPCS pump and p.,

motor after it had been turned over for preoperational

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Contrary to the requirements of SU.ACP.16 above, on L ;'

Earch 17, 1979,'the licensee flooded the RBR-A/LPCS E

(M pump room submerging the system jockey pump and motor, I-d J -~

instrumentation and valves of both systems, of which h--j d LPCS system had been turned over for preoperational c2 - J testing.

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NRC Form 310B (R ZZI)(3 78) NRCM O240 e a manwtut paints wagt, a,7e.asm?

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Appendix A >

b.

Step 5.1.5 of the operating procedure OP.HP.01-4, Revi-

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sion 0 for lineup, fill and vent of the BPCS system requires that the operator close valves IE22-F003 and IE22-F031.

j i Contrary to the above, on January 19, 1979, when the HPCS system was started, because valves 1E22-F003 and 7031 had been left open, high pressure core spray water entered the condensate and low pressure core spray system causing a rupture of the steam jet air ejector l

condensor 1A.

c.

Switching Order No. 781317, dated November 16, 1978, re-quired valve 1E21-F025 to be safety tagged and closed.

This valve completes a crossconnection of the LPCS and HPCS systems.

Contrary to the above, on January 19, 1979, the valve IR21-F025 was in the open position when the HPCS system was started resulting in overpressurization of the LPCS system piping.

2.

10 CFR Part 50 Appendix B, Criterion V, states in part that,

" Activities affecting quality shall be prescribed by documented instructions, procedures or drawings, of a type appropriate c.

to the circumstances...." Paragraph 17.1.5.1 of the FSAR states in part that, " Activities affecting the quality of the facility are accomplished in accordance with written instructions, procedures or drawings which prescribe accep-table methods for carrying out the activities...."

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, ' 71, Contrary to the above, the inspection Procedure 22A4387 E'",'r b Revision 4 and the Field Deviation Disposition Report No.

Ec ra RN-1-286 used for inspection of the control rod blades by Q3 GQ the licensee during the months of July through October of Od 1 E_s 1978 did not specify what readings were desired, where e} dr>)

such readings were to be e=kan on the blade and specifically, cp how the r1==p was to be used.

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10 CFR Part 50 Appendix B, Criterion IVII, states in part that.

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" Sufficient records shall be amintained to furnish evidence of activities affecting quality.

... Inspection and test records shall, as a =fn4=r, identify the inspector or data recorder, the type of observation, the results, the acceptability and the action taken in connection with any deficiencies noted."

Section 17.1.17 of the FSAR requires that sufficient records en fm nt.h evid-nee of merivief aa af fectine min 14 ev be

- maintained a t the site or available fcr audit. Records ftff TAmpect1nor and tustr-inulada; 1ormor-ltaired-w-SURNAME)

D^ N NRC Form 318s (RZn)(178) NRCM 0240

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1 Appendix A inspector or data recorder, type of observation, results, ecceptability and results of deficiencies.

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Contrary to the above, the licensee's contractor failed to maintain sufficient records for the inspection of control rod blades in accordance with the Field Deviation Report (FDDR) No. EN-1-286.

No raw data existed from the inspection to verify the conclusions presented by the FDDR.

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l U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION III Report No. 50-358/79-06 Docket No. 50-358 License No. CPPR-88 Licensee:

Cincinnati Gas and Electric Company 139 East 4th Street Cincinnati, OH Facility Name:

Wm. H. Zimmer Nuclear Power Station Inspection At:

Wm. H. Zimmer Site, Moscow, Ohio Inspection Conducted: February 27-28, March 1-2, 19-23, and April 9-11, 1979 4

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Inspectors:

1. w~ 4 R. W. Dettenmeier fM/79 fdw/

kOl['79 B. M. k. Wong ( ril 9-11)

Approved By:

J. F. Str qter, Chi Nuclear S port Se ion 1 Inspection Summary Inspection on February 27-28, March 1-2, 19-23, and April 9-11, 1979 (Report No. 50-358/79-06)

Areas Inspected:

Routine, unannounced inspection of Preoperational Test Program; previous unresolved items and actions on previous inspec-tion findings; preoperational test results; system turnover for pre-operational testing; control rod blade inspection; licensee events.

The inspection involved 160 inspector-hours onsite by three NRC inspectors.

Results: Of the six areas inspected, no items of noncompliance or deviations were identified in four areas. Three items of noncom-pliance (two infractions - failure to follow procedures - Paragraphs 7, 8.b, 8.c, 9.c.; inadequate procedure - Paragraph 10.f.; and one deficiency - inadequate records - Paragraph 10.e) were identified in two areas.

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DETAILS 1.

Persons Contacted

• J. Schott, Station Superintendent

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P. King, Assistant Station Superintendent S. Martin, Test Coordinator R. Ldnk, Operations Supervisor R. Price, Training Supervisor D. Anderson, Turnover Coordinator

• J. Wald, Station Quality Engineer

• • • B. Culver, Project Manager

• • W. Schwiers, Principal Quality Assurance and Standards Engineer

• • • R. Wood, QA&S Engineer

• H. Gear, Construction Supervisor

• • • T. Bloom, GE Resident Site Manager M. May, GE Site Operations Manager

• • • L. Aiello, GE Quality Assurance Engineer V. Pence, GE Unit Manager, Driveline Components J. Occoner, Reactor Controls, Inc.

The inspectors also interviewed otter licensee employees including members of the administrative, technical, operating and QA&S staff; employees of the General Electric Company, employees of EDS Nuclear, and employees of Reactor Controls, Incorporated.

• Denotes those attending the exit interview of March 23, 1979.

• • Denotes those attending the exit inter.'iew of March 23 and April'11, 1979.

• • • Denotes those attending the exit interview of April 11, 1979.

2.

Licensee Action on Previous Inspection Findings (Closed) Noncompliance (358/79-01-01).

Failure to follow safety tagging (switching order) procedure.

The inspector found that the licensee is conducting safety tagging refresher training for all operations personnel and systems engineers as stated in their letter, Borgmann to Heishman, dated February 28, 1979.

(Closed) Noncompliance (358/79-01-04).

Failure to develop appro-priate procedure to implement to QA&S responsibility assigned in SU.ACP.03.

The inspector found that the QA&S Principal Engineer has issued a letter to the Station Superintendent requesting QA&S be informed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> in advance of a flush of a safety re-lated system so that appropriate audits can be conducted on the valve lineup. A QA&S engineer has been assigned responsibility to perform the audits. According to EPD personnel the responsibi-lity to inform QA&S has been delegated to the system engineer in charge of the flush. This action is in accordance with the response given in the letter, Borgmann to Heishman, dated February 28, 1979.,o L *) i

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r (Open) Unresolved Item (358/78-03-02).

Three core support plate pins " repaired" prior to taking bend measurements. During a meeting between the licensee, General Electric Company personnel, and the inspectors, the problem of what cculd happen if the three pins failed in time, due to the unknown level of cold working each pin experienced, and how to monitor for such failu,re was discussed.

No conclusions were reached.

The inspector stated that the licensee's written response to the NRC on this subject to close out the outstanding 50.55(e) report should address the following items:

State why the licensee feels the pins will not fail in time a.

(should not be replaced) and give the basis for such position.

b.

Assume the pins will fail and state what possible effects, if any, it could have on plant operation, refueling, mainte-nance on control rod blades, etc.

Among the possible effects, consider possible rotation of the fuel support piece so that the coolant flow paths do not align with the path in the control rod guide tube.

Discuss what methods are available, if any, to monitor for c.

the possible failure of the pins.

Include among the methods visual examination during each refueling outage.

List the methods the licensee will commit to follow.

d.

Commit to develop prior to power operation the acceptance criteria, for each of the monitoring programs committed to in Item c., and the required corrective action whenever one of the limits is exceeded.

(Closed) Unresolved Item (358/79-01-02).

Operator training on the performance of valve check lists and switching orders.

The inspector found that the licensee is revising Station Administrative Directive (SAD) OS. SAD.02, Revision 1, " Station Operations" to include under Paragraph 5.1.7 the specific criteria to be followed by plant operators during the performance of checklists.

Operators become familiar with changes to SAD's through the use of the All Read Folder which they are required to read once/ month and sign off.

In addition, the licensee stated all Shift Supervisors are required to discuss the changes with his crew.

Regarding operator performance of switching orders, the licensee has written an Operations Memo on the subject which requires the operator to place the control switch in the close or open position, whichever is called for in the order, to verify the valve position.

3.

Other Areas Inspected (Closed) 10 CFR 50.55(e) Event No. M-ll, Control Rod Interferences.

The inspectors found that the licensee had modified 80 of the control rods by grirding a chamfer in the upper corner of the 2 U7/

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o control rod velocity limiter.

The licensec completed this modification on site in accordance with the General Electric Field Dispositiod Instruction No. 94/6300, Revision 0 to elminate the potential interferences between the control rod and t'se fuel channel. A review of the control rod inspection records for the 80 control rods modified revealed that:

51 control rods were modified on all foar corners.

10 control rods were modified on thtee corners.

11 control rods were modified on tso corners.

8 control rods were modified on one ccrner.

No items of noncompliance or deviatior.s were identified.

4.

Preoperational Test Program The inspector reviewed the revisior to the Startup Administrative Control Procedures (ACP's) and Sta tup Project Procedures as of March 2, 1970, to ensure the chanses do not conflict with FSAR commitments. The inspector stated that ACP No. 18, " Work Requests" is considered unacceptable and must be rewritten as a Station Administrative Directive in line with the comments given to the licensee by the Project Inspector.

As of April 4, 1979, of the 114 preoperational tests required to be completed prior to fuel loading the licensee has completed writing 108 and has approved for use 77 test procedures.

Twenty-two systems or partial systems have been turned over for preopera-tional testing, 17 preops are in progress and four tests have been completed. No test results have been approved by the SRB yet.

No items of noncompliance or deviations were identified.

5.

Preoperational Test Results The inspectors reviewed the results of the testing performed as of March 20, 1979, on the 24/48 VDC and 125 VDC systems.

During the review it was noted that:

Stratification of the electrolyte in the 24/48 VDC lead-calcium a.

battery cells is creating an operational problem for the licensee during recharge following a disebarge test in that it took approximately two months before the specific gravity of the sample obtained at the top of the cells reached an acceptable reading ( l.205).

While most lead-calcium cells have an electrolyte withdrawal tube which permits sampling at a point one-third down from the top of the plates and, according to the vendor, gives a more accurate indication of the state of charge, the Model DC-9 being used for the 24/48 VDC system is not equipped with such a tube.

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The inspector stated that while the test results meet the acceptance criteria, the length of time it took to obtain acceptable specific gravity readings appears to be unde-sirable during plant operation.

Some method to obtain a more representative sample of the electrolyte should be considered.

b.

Two cells on the IB 125 VDC battery failed to meet the acceptance criteria (cell #84 specific gravity of 1.201, and cell #88 voltage of 2.06 volts).

The licensee plans to individually recharge these two cells prior to the next test which is the 80% of one-hour rating discharge test.

The inspectors stated that if these two cells fail to meet the acceptance criteria following the next discharge test, they should be replaced.

This is an unresolved item (358/79-06-01) pending the results of the next discharge test.

c.

Specific gravity readings were informally corrected by the licensee for changes in the electrolyte level without affecting the results of the tests. The licensee has agreed to require such corrections as part of his test procedure.

No items of noncompliance or deviations were identified.

6.

Review of System Turnover for Preoperational Testing The inspectors reviewed the system turnover process for system release for preoperational testing to ensure compliance with the Startup Manual procedures. The review revealed that:

a.

The Master Punchlist, which consolidates the lists of items remaining to be completed, although continuously updated by the turnover group, appeared to be missing an approximately month old item for the low pressure core spray release pack-age identified by one of the licensee's constructica con-tractors. Members of the licensee's staff indicated that a problem existed in the adequacy of the continuous updating of the Master Punchlist and that this inadequacy was impacting on the performance of the preoperational testing.

The licensee indicated that the system for compilation of the Master Punchlist was only recently enacted and that the numerous sources of input to the system created problems in assurring a complete punchlist.

The licensee has initiated a program to compile punchlist items on computer printout listings as well as to define responsibilities among the licensee's staff. This is an unresolved item (358/79-06-02) pending further review of the system turnover for preoperational testing by the inspectors. i ioj

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b.

When revisions are made to drawings on systems which inter-connect or interface with the scoped system for release for preoperational testing, and especially in the case of electrical drawings where the connecting drawings are part of the release boundary, the system engineers stated that they felt they were not given enough information in the revision notification to determine how the revision would'effect or impact upon the preoperational test or results.

It was indicated by the licensee that although additional research might be involved on the part of the system engineer, all information for determining the impact of revisions to interconnecting or interfacing drawings to the preoperational test or its results is available somewhere on site.

It was also indicated by the licensee that the turnover group will prepare a brief summary of how the revisions to drawings of interconnecting or interfacing systems will effect the scoped system for release.

This problem with interconnecting and interfacing systems originates from the fact that there is a difference in the number of drawings used in preparing the preoperational test and the number of drawings used when scoping the boundaries of the system for release for preoperational testing.

Although the drawings contained in the release package are frozen to the point that no new revisions of the drawings will be released without authorization from the system engineer, the system engineer has no control of the release of revisions of drawings which were used in preparation of the preoperational test procedure which are not also inclu-ded in the release package for turnover for preoperational testing. The inspectors indicated that these differences in capability to control revisions to drawings which could effect the preoperational test could lead to portions of systems or interconnecting portions between systems not being preoperationally tested. This is an unresolved item (358/79-06-03) pending further review of the preoperational test program by the inspectors.

There seems to be some confusion among the licensee's staff c.

members as to a method for dealing with revisions to drawings for systems which have been preoperationally tested but not turned over for operation. The licensee indicated that procedures are already developed to handle revisions as well as how the revisions affect the validity of the preoperational test results. The licensee indicated that he will review these procedures with the staff personnel.

d.

The licensee has procedures for control and tagging of areas, com onents, panels and systems which have been turned e

over for preoperational testing. These procedures 02?

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do not address situations where components or panels are shared by one system which is turned over and by another which is not.

During the plant tour, it was found that the

" Division I LPCS/RHR-A Relay Board" did not have any of its internal components tagged as being turned over even though the LPCS system had been turned over.

The licensee indicated that tagging the panel or all of the LPCS relay components within the panel or any other similar situation was a matter of judgement since construction was still allowed on the RHR-A portion of the panel.

It was also noted during the plant tour that most of the electrical control and relay panels such as the "HPCS Relay Board" which was tagged as turned over had their doors completely removed even though there was no work being conducted within the panels in most cases.

The inspectors questioned the ability to control further work by construc-tion in such areas as well as in shared system panels.

The licensee indicated that the procedures for " System Release for Turnover" SU.RPR.01 and " Construction Work Authorization for Equipment Turned Ov for Preoperational Testing",

SU.PRP.04 controlled construction work on and around turned over systems, components and areas.

The licensee indicated that QA&S also audits the proper completion of the authori-zation forms involved.

Field audits to verify conformance to the above mentioned procedures to verify actual initiation of authorization forms are not done.

The licensee indicated that he would review the above procedures with regard en the inspectors comments.

This is an unresolved item (358/

79-06-04) pending resolution by the licensee and further review by the inspectors.

In discussions with the system engineers, turnover groups e.

and operations personnel it was indicated that there was some confusion as to the disposition of the " Construction Work Authorization" form PRP-04-1 and the " Return of System /

Equipment to Construction" startup form 6.5 while construc-tion work was in progress under this authorization.

The operations personnel indicated that some of the above men-tioned forms were available in the control room although there was no way to know if they reflected all work going on under all such authorizations since there was no designated disposition while work was in progress nor a sequential numbering system established for filing or traceability. The licensee indicated that operations personnel have a definite need for notification of such authorization and would resolve the questions concerning the authorization forms. This is an unresolved item (358/79-06-05) pending resolution by the licensee and further review by the inspectors.

No items of noncompliance or deviations were identified. 2h

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Plant Tour The inspectors conducted tours of various areas of the plant to observe activities in progress, general housekeeping and clean-liness, and equipment caution safety and or green preoperational testing tagging. The tours revealed that:

The enclosed area of the spare and IB 125VDC battery chargers, a.

which have been turned over for preoperational testing, was cluttered with debris. Brooms, general trash, metal grating, hard hats and a welding oven (PL-5) surrounded the IB battery charger within the wooden barricade.

b.

The spare 125VDC battery charger had one of its side panels missing. Pieces of insulation and concrete chips and heavy dust deposits had settled onto the internals of the spare charger.

It was noted by the licensee's staff that the panel had been missing at least since the chargers had been turned over for preoperational testing in October 1978.

Startup Manual procedure SU.ACP.16 " Equipment and Building Cleanliness" requires in Section 5.2 that all safety related equipment shall be maintained free of excessive oil, water or other material which could prevent the equipment from performing its intended safety function. The licensee has failed to follow the Startup Manual procedure SU.ACP.16.

This failure to follow procedures is contrary to the requirements of 10 CFR 50 Appendix B, Criterion V, and is considered to be an example of an item of noncompliance (358/79-06-06A) of the infraction level.

8.

Overpressurization of Low Pressure Core Spray and Condensate Systems Piping The inspector reviewed the event of January 19, 1979, during which high pressure core spray (HPCS) water entered the conden-sate (CD) and low pressure core spray (LPCS) systems because valves 1E22-F003 and F031 had been left open causing a rupture of the steam jet air ejector condenser IA.

The review consisted of interviews with testing and operating personnel and a review of the licensee's final report on his investigation of the event.

The review showed that:

Procedure OP.HP.01-4, Revision 0 was used to lineup, fill a.

and vent the HPCS system.

b.

At the completion of the fill and vent operation the operator never completed Step 5.1.5 which required him to close valves lE22-F003 and F-31.

With these two valves open the CD and HPCS systems became crosstied thru the cycled condensate (CY) system. The operator claims he informed the Shift 5

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s Supervisor that he had left the two valves open while the latter does not recall being told.

This failure to follow procedures is contrary to 10 CFR Part 50, Appendix B, Criterion V and is considered to be an example of an item of noncompliance (358/79-06-06B) of the infraction level.

For some unknown reason, valve IE21-F025 which ha been c.

safety tagged closed under Switching Order No. 781317, dated November 16, 1978, was in the open position.

This completed the cross connection of the LPCS and HPCS systems. Violation of Switching Order No. 781317 is contrary to 10 CFR 50, Appendix B, Criterion V and is considered an example of an item of noncompliance -(358-79-06-06C) of the infraction level.

The switching order was cleared on January 24, 1979.

The corrective action which the licensee is currently taking regarding a previous noncompliance with the safety tagging procedure (358-79-01-01) is also applicable to this event, therefore the inspector stated no response to this item of noncompliance is required.

d.

Paragraph 13.0 of Jafety Tagging Procedure EC. SAD.02, Revision 00 allows for the operation of equipment for test purposes without the removal of the safety tags.

It is possible that valve 1E21-F025 was operated for test purposes thru tags and subsequently left open by error.

The inspectors have objected to Paragraph 13.0 of the Safety Tagging Procedure.

On March 21, 1979, the licensee issued operating memo 79-2, Revision 9, which specifically requires that "Do Not Operate" tags must be removed before energizing electrical equipment or opening valves. An exception is made in the case of electrical testing conducted by EOTD in which case only the EOTD master tag will be left in place.

e.

On December 12, 1977, a General Electric system engineer recommended that a check valve be installed on line 1HP18A3 downstream of valve IE22-F031 because a similar overpres-surization of a small section (up to valve ICY 013) of low pressure piping had occurred.

The recommendation was re-jected because the licensee thought that two valves (IE22-F003 and F031) plus adminstrative controls were sufficient to prevent recurrence.

The licensee stated the check valve will be installed. All other ECCS systems have check valves in the line from the CY system.

The inspector stated his concern regarding repeatable occur-rences where a lack of communication or understanding between parties have resulted in damage to equipment.

It is our intention to closely monitor the licensee's performance during the preoperational test program to determine the U / /p l4b

I

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the adequacy of plant staffing and training as fuel load date approaches.

9.

Flooding of LPCS/RHR-A Pump Room Event The inspectors reviewed the event which occurred on March 17, 1979, involving the LPCS/RHR-A pump room.

The inspectors inter-viewed construction and operations personnel involved in the event and reviewed the operations and shift engineering logs.

The review revealed that:

a.

On March 17, 1979, flushing was being conducted on the fuel pool cooling and cleanup system.

flushes to the suppression pool.

The licensee normally unavailtble due to modifications. The suppression pool was This flush was to the clean-up phase separator tank which was lined up to the reactor building equipment drain tank for overflow.

b.

At lunchtime the engineer in charge of the flush told the pipe fitters working under him to close the valves to secure flushing and break for lunch.

The valves were not closed to stop the flush.

Water overflowed from the reactor building equipment drain tank and into the corner room which contained the RHR-A, LPCS and jockey pumps and its associated instruments and valves for the LPCS system as well as the RHR-A system.

The configuration of the corner room is such that it will act as a pool up to approximately five feet.

When water was discovered flowing out of this corner room into the annular the base of the wetwell the uncontrolled flush was area at terminated by closure of the valves from the condensate storage tank supply.

The sump pumps in the RHR-A and LPCS pump room had been out of service due to maintenance.

LPCS system has been turned over for preoperational testing.

The licensee indicated that all instrumentation involved in the flood of the RHR-A/LPCS pump room would be dried inspected and recalibrated.

, cleaned Restoration of the instrumen-tation had already begun while the inspectors were on site The licensee also indicated that the method of using available storage tanks for deposition of flushing waters would not be continued due to the problems experienced.

The licensee did however, indicate what methods would be used during the

not, time that the suppression pool was not available.

This event, as well as others experienced by the licensee in recent months are considered to be examples of poor com=unications experienced during operation of systems for testing and flushing.

The inspectors indicated to the licensee that a need for developing better communications existed.

This is considered to be an unresolved e [y

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item (358/79-06-07) pending resolution by the licensee and further review by the inspectors.

During the flushing on the 17th of March, 1979, the floor c.

drains were also filled and backedup in the reactor building at the 503 level.

This backup resulted in a water washdown of the HPCS motor and pump.

It was indicated by,the operating staff that the HPCS pump motor and the RHR-C pump motor, both in the same corner room, had been washed down in similar events several times in the past few months. HPCS has been turned over for preoperational testing.

Startup Manual Procedures SU.ACP.16 " Equipment and Building Cleanliness" requires in Section 5.2 that all safety related equipment shall be maintained free of excessive oil, water or other material which could prevent the equipment from performing its intended safety functions.

The licensee has failed to follow Startup Manual Procedure SU.ACP.16 in the cases of the flooding of the LPCS/RHR-A pump room and the wash down of the HPCS pump motor. This failure to follow procedures is contrary to the requirements of 10 CFR 50 Appendix B, Criterion V. These two failures to follow procedures are considered examples of items of noncompliance (358/79-06-06D) of the infraction level.

10.

Control Rod Elades Inspection The inspectors reviewed the results of the control rod blade inspections performed by the licensee during the months of July thru October 1978.

Our review revealed that:

Of the 137 control rod blades originally inspected using a.

0.280 inch and 0.320 inch envelope gauges, 86 failed to pass the.280 inch envelope gauge and of those 86, four also failed the.320 inch gauge.

b.

In accordance with the Inspection Procedure 22A4387, a clamp, which applied approximately 40 pounds of pressure against the blade sheath, was placed approximately one to two inches from the area in question (it should be noted that the procedure does not state where to place the clamp).

Thickness measuremet.ts were taken with a micrometer and of the 86 control rod blades 11 still exceeded.280 inches in thickness. According to Reactor Controls, Inc. (RCI) records the 11 blades in question were:

C. R. Blade Serial No.

Blade No.

High Spot A400 1

0.286 A420 4

0.183

'A435 2

0.289.</

I

A440 2

0.281 A443 2

0.282 A453 1

0.281 A461 3

0.285 A484 3

0.285 A501 2

0.282 A510 2

0.290 :

A515 4

0.282 Of the 11 control rod blades four (A400, A435, A443 and c.

A461) also failed the 0.320 inch gauge and were rejected per Field Deviation Disposition Request (FDDR) No. KN-1-288.

Blade A484 was rejected solely on the basis of a 0.285 inch reading. The exact location of the high area was not re-corded.

In addition, control rod blade A437 was rejected on the same FDDR due to numerous nicks in the sealing area.

d.

The Inspection Procedure 22A4387 R4 was modified and the changes recorded by FDDR No. KN-1-286.

On the FDDR the exact size and location of the high area was determined for the remaining six control rod blades (A420, A440, A453, A501, A510 and A515).

The size of the inspection area was reduced and the 40 pound clamp was placed directly over the high point area while deep throat micrometer readings were taken around the clamped area.

As a result of FDDR No. KN-1-286 the six blades in question e.

were accepted on the basis that the thickness of the high point area was reduced to less than 0.280 on five of the blades with the 40 pound clamp over the area.

For the sixth blade (A-510) the thickness remained at 0.290, but the blade was accepted on the basis that half of the high point area was located outside the reduced area of interest and a statement that it " met the intent" of 22A4387.

A review of RCI's records showed that no records were kept of the work done as a result of FDDR No. KN-1-286.

No raw data exists to verify the conclusions presented by the FDDR.

Failure to maintain sufficient records is contrary to 10 CFR 50 Appendix B, Criterion XVII and Paragraph 17.1.17 of the FSAR and is considered to be an item of noncompliance (358/79-06-08) of the deficiency level.

On April 10, 1979, the air. blades in question were reinspec-ted by the licensee, RCI,...; the General Electric Company.

The reinspection was witnessed by NRC inspectors. Accurate, detailed records of the reinspection were kept this time 9/5 148

(

and as a result no response is required from the licensee on the above item of noncompliance.

f.

During the review of Inspection Procedure 22A4387 and FDDR No. KN-1-286, which supposedly clarified the inspection procedure, it was noted that both failed to rpecify what direct readings were desired, where such readings were to be taken, how the clamp was to be used, etc. and as a result, both the Inspection Procedure and the FDDR are considered to be inadequate. This inadequacy is contrary to 10 CFR 50 Appendix B, Criterion V and Paragraph 17.1.5 of the FSAR and is considered to be an example of an item of noncompliance (358/79-06-09) of the infraction level.

A meeting was held on April 10, 1979, between the General Electric control rod design engineer, the licensee and the NRC inspectors at which time the blade acceptance criteria

(.280"), the use of the 40 pound clamp and the results of thin control rod qualification tests, etc. were discussed.

The prototype tests were performed by the General Electric Company using 0.080" and 0.120" vall fuel channels for the purpose of determining the degree of misalignment, channel deformation, water gap reduction, etc. at which the opera-tional performance of the control blade would be affected.

The test results showed that considerable misalignment (between 10 to 14 times the allowable during core internals installation); or reduction of the water gap, due to mis-alignment and channel deformation, by a factor of approxi-mately 1.5 to 2.0 would be required before operational dif-ficulties were first experienced. Wear of the control rod sheath and fuel channels was measured on tests conducted for the designed 20 year life cycle. Based on the results of these tests and of the reinspection performed on April 10, 1979, the control rod blades presently supplied to the Zimmer Station are considered to be satisfactory.

During the discussion with the General Electric company personnel it was stated that for BWR's 5 it is recommended that the fuel channels be rotated during each refueling outage.

Since this is new information and since it was not clear how and when the channels are to be rotated the inspector requested that the licensee obtain written clarification of General Electric's position regarding the desirability or need for channel rotation.

This is an unresolved item (358/79-06-10) pending further review of this matter by the licensee and the inspector. i' r I l il )

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11.

Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, items of noncompliance, or deviations.

Unresolved items dis-closed during the inspection are discussed in Paragraph _s 5.b, 6.a. 6.b, 6.d, 6.e. 9.b, and 10.

12.

Exit Interview The inspectors met with licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection on March 23, and April 11, 1979.

The inspectors summarized the purpose and the scope of the inspection and the findings.

In response to certain of the items discussed by the inspector, the licensee representatives:

Acknowledged the statements by the inspector with respect to the items of noncompliance (Paragraph 7, 8, 9 and 10).

Stated the SAD covering " Work Requests" would be completed by March 31, 1979 (Paragraph 4).

Objected to the inspector's position regarding the three core support plate pins (Paragraph 2).

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APR 9 - 1979 Docket No. 50-358 '

Cincinnati Gas and Electric Company ATIN: Mr. Earl A. Borgacan Vice President Engineering Services and Electric Prodnetion 139 East 4th Street ci n, 4,== ti, OH 45201 Centlem n:

This refers to the inspection conducted by Hessrs. J. Hughes, T. E. Vandel and H. H. Wescott of this office on March 21-23, 1979, of activities at the un H. Zimmer Nuclear Power Station authorized by NRC Constructim Pemit No. CPPR-88 and to the discussion of our findings with Mr. B. E. Culver and others of your staff at the conclusion of the inspection.

The enclosed copy of our > 'spection report identifies areas i.

• v= inad during the inspection. Within these areas, the inspection consisted of a selective --==ination of procedures and representative records, observations, and interviews with personnel.

No items of noncourp11mnee with HRC requirements were identified during the course of this inspection.

In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and the enclosed inspection report will be placed in the NRC's Public Document Room, except as follows.

If this report contains information that you or your contractors believe to be propriatary, you must apply in writing to this office, within twenty days of your receipt of this letter, to withhold such information from public disclosure. The application must include a full statement of the r===aan for which the information is considered proprietary, and should be prepared so that proprietary information identified in the application is conemin.A in an enclosure to the application.

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Cincinnat1 Cas and.

Electric Company APR 9 - B79 W will gladly discuss any questions you have concerning this inspection.

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Sincerely, C. Fiore111,(11.F Reactor Construction and Engineering Support Eranch

Enclosure:

IE Inspection R* Port No. 50-358/79-09 cc w/ enc 1:

J. R. Schott, Plant Superintendant Central Files Reproduction Unit NRC 20b PDR Local PDR ESIC TIC U. Young Park, Power Siting Connission Citizens Against a Radioactive Environment e4 --

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xt U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT REGION III Reporr.4o. 50-358/79-09 Docket No. 50-358 License No. CPPR-88 Licensee: Cincinnati Gas and Electric Company 139 East 4th Street Cincinnati, OH 45201 Facility Name: Wm H. Zimmer Nuclear Power Station Inspection At:

Zimmer Site, Moscow, Ohio Inspection Conducte : Mar 21-23, 1979 fth l

Inspectors:

T. E. Vandel

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H. M. Wescott

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Approved By:

. C. Rnop, Chief

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Projects Section

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Inspection Summary Inspection on March 21-23, 1979 (Report No. 50-358/79-09)

Areas Inspected:

Follow up of previously identified noncompliance and unresolved matters; review of electrical and instrumentation activities in progress and related records; RER system as built walk down; Part 50 Section 50.55(e) and Part 21 reports review. The inspection involved a total of 66 inspector-hours onsite by three NRC inspectors.

Results:

Of the seven areas inspected, no items of noncompliance or deviations were identified.

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DETAILS Persons Contacted Cincinnati Gas and Electric Company (CG&E)

• E. K. Culver, Project Manager

• R. P. Ehas, QA&S Engineer wJ. W. Haff, QA&S Engineer

• J. F. Weissenberg, QA&S Engineer W. W. Schwiers, Principal QA&S Engineer D. Kramer, QA&S Engineer R. L. Wood, QA&S Engineer Henry J. Kaiser Company (Kaiser)

• R. E. Turner, QA Manager R. Marshall, Project Manager J. T. Setlock, Senior QA Mechanical Inspector

• Denotes those personnel attending the exit interview held at the end of the inspection (March 23, 1979).

Other personnel of CG&E and Kaiser were contacted during the course of the inspection.

Licensee Action on Previous Inspection Findings (Closed) Unresolved Item (358/78-13-01):

Several welds on seismic Class 1 cable tray hanger components in the reactor building cable areas were of questionable quality relative to the requirements of AWS DI.1-72.

Repair of the defective welds are in process, the RIII inspector observed several completed welds and determined that the correct Kaiser velding procedure, SPP, 3.1.11 and S&L Seismic Category I Electrical Equipment Hangers Fabrication and Erection Specifications, STD-ED-115 w3re being used. Kaiser QC inspectors are inspecting all weld repairs and documenting on drawing E-19 and nonconformance report E1139 Revision. This item will continue to be inspected as a part of the Reportable Event 10 CFR 50.55(e) dated July 17, 1978.

(Closed) Unresolved Item (358/78-29-02):

Tech-Sil, Incorporated Quality Assurance Program. CG&E's QA conducting an audit on Tech-Sil, Incorporated corporate office. The RIII inspector reviewed CG&E's Field Audit Report No. 219 conducted on the L:plementation of Tech-Sil, Incorporated QA program. The RIII inspector has no further questions at this time.

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(Closed) Noncompliance (358/78-21-01; 02):

These items of nonco=pliance were closed in inspection report 358/79-03. However, there remained a question as to what action would be taken to preclude repetition of like items by other site subcontractors to resolved any generic question.

The inspector reviewed three (3) licensee " Field Audit Reports" as follows:

1.

No. 222, dated February 20, 1979, audit of Reactor Controls incorporated.

2.

No. 224, dated March 2, 1979, audit of Henry J. Kaiser Company.

3.

No. 226, dated March 13, 1979, audit of Waldinger, Young and Bertke.

The above appeared to be satisfactory and requires no further action by the inspector.

(Closed) Noncompliance (358/79-07-06):

Weld rod hold ovens were not equipped with direct read out temperature indicators although these ovens indicated that they had been calibrated on September 20, 1978.

The inspector verified that the veld rod hold ovens No. W-12, W-49 and W-52 had been removed from service as stated in the licensee response letter.

Other Inspection Areas 1.

Licensee Actions Pursuant to 10 CFR 21 The Jamesbury Corporation, supplier of mounting brackets (four),

for certain valve actuators reported pursuant to 10 CFR 21 that the brackets did not meet specified seismic requirements.

During the current inspection the following documents were reviewed relative to this matter:

(1) licensee's nonconformance report (NR) dated January 23, 1979, which identifies the pertinent 8" butterfly valves (1WS037A, B, C and D); (2) Acton Environmental Testing Laboratories seismic test report No. 13865, Revision 1, dated October 16, 1978; (3) Jamesbury Corporation certification for addendum "B" of report No. JHA-74-2 dated November 20, 1978; and (4) Sargent and Lundys's (A/E) specification B-22 for valves.

During his review, the inspector noted that the orginally specified seismic loads on the subject valves had been changed from SG to 3G.

The licensee explained to the inspector that the A/E had redone a previous calculation and established that the acceleration loading criteria for these valves is 1.5G.

Based on this, the loading requirement was reduced from SG to 3G. $!'L k

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i Review of the seismic test report for these Jamesbury valve actuators and mounting brackets indicates that, in each case, brackets attached to the valves, vibrated loose. The inspector was unable to determine if this test was in fact a valid test.

These items will be reviewed during a future inspection.

2.

10 CFR 50.55(e) Reports a.

480 Volt Electric Operated Circuit Breaker The inspector visited the licensee maintenance shop to observe a circuit breaker of the type identified in the licensee defi-ciency report. The breaker involved is an ITE electrically operated 480 volt breaker type K6005.

It was observed that the trip bar shaft has sufficient horizontal movement tolerance that if movement occured all in one direction, toward the trip i

solenoid end, the latch lever, that latches the breaker in the open position, binds'on the solenoid and is prevented from moving freely to latch out the breaker. The breaker then is free to reclose when the closing spring becomes fully charged. The manufacturer is reviewing alternatives for a satisfactory corrective action to prevent future undesired closings. Review will be conducted during future inspections.

b.

Inadequate Welding of Seismic Class 1 Cable Tray Hangers This item, was originally identified as an unresolved item, and subsequently was reported in accordance with Part 50, Section 50.55(e) requirements relative to supports provided by Superstrut Incorporated. Disposition of the matter was resolved earlier in this report. Complete close out will involve the review of the licensee's action regarding the repairs of the hanger welding as outlined in their final report dated October 30, 1978. V, m k{

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Section 1 Prepared by H. M. Wescott Reviewed by R. C. Knop, Chief Projects Section 1.

Review of Quality Records for Safety Related Piping, Residual Heat Removal (RHR), Welding The inspector reviewed quality records related to welding of safety related piping outside the reactor coolant pressure boundary to ascertain that these records reflect work accomplishment consis-tent with NRC requirements and SAR commitments as follows:

Kaiser Engineers, Incorporated, Field Weld Data Sheets for the a.

Wm H. Zimmer Nuclear Power Station as follows:

(1) FW No. 76RH, Line No. 1RH07BA10 (2) FW No. 58RH, Line No. IRH06AA14 (3) FW No. 61RH, Line No. 1RH05AA14 These field welds were selected at random during walk down of RHR system.

The records contained Weld Material Issue slips, Radiography Reports, Exposure Technique sheets and Liquid Penetrant inspection Reports.

b.

Subassembly Packages from Pullman Kellog for pipe spools as follows:

(1) Spool No. IRH55BB10-261. This package contained Defective Material Report No. 4523, spool replaced on P0 17-01-00-75-33K, dated September 29, 1979.

(2) Spool IRH-55BB12-262 (3) Spool IRH-55BB12-263 (4) Spool IRH-55BB12-264 (5) Spool IRH-55BB12-265 These packages contained certificates of Shop Inspection, Weld History Records, Radiographic Reports, Magnetic Particle Inspection Records, Demensional Inspection Records and Final Inspection Checklists.

No items of noncompliance or deviations were identified. ;

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2.

Reactor Coolant Pressure Boundary Piping (Welding) Observation of Work Residual Heat Removal (RHR) System The inspector reviewed and made observations of completed work of the RER system to ascertain that the system was installed in accordance with NRC requirements as follows:

a.

Review of Wb H. Zimmer Drawings of the RER system as'.follows:

(1)

M-51, sheet 1, Revision R, dated August 10, 1978.

(2)

M-51, sheet 2 Revision P, dated August 10, 1978.

(3)

M-51, sheet 3, Revision L, dated Janaury 6, 1978.

(4)

M-51, sheet 4, Revision N dated. November 17, 1978.

b.

The inspector preformed a walk down of the RHR system to ascertain that the equipment was installed as required.

It was established that no block valves had been installed inside the containment on the suction side of the RHR pumps A, B and C as required by 10 CFR 50, Appendix A

" General Design Criteria for Nuclear Power Plants" Criterion 56 - Primary containment isolation, which requires that two isolation valves be installed in lines penetrating the containment (one inside containment and one outside of containment installed as close as practical to the containment).

Review of the Safety Evaluation Report and discussion with NRC licensing established that the requirement for isolation valves inside of containment had been vaived as it could be demonstrated that they were not required.

No items of noncompliance or deviations were identified.

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Section II Prepared by J. Hughes Reviewed by D. W. Hayes, Chief Engineering Support Section 1 1.

Instrumentation - Observation of Work Activities a.

The inspector observed four installed Class IE instrument piping sensing lines identified by the following instruments and KII isometric drawings:

(1) Reactor vessel pressure transmitter:

Drawings - M471-21-NB-298, Revision 3 M471-21-NB-296, Revision 2 M471-1-NB-13, Revision 9 M471-1-NB-16, Revision 5 (2) Reactor drywell pressure switch:

Drawing - M471-18-NB-354, Revision (3) RHR discharge pressure switch:

Drawings - M471-10-RH-41, Revision 10 M471-10-RH-50, Revision 14 M471-10-RH-51, Revision 4 (4) RER pressure differential indicating switch:

Drawings - M471-10-RH-32, Revision 14 M471-10-RH-33 Revision 11 M471-10-RH-34, Revision 5 M471-10-RH-35, Revision 15 M471-10-RH-36, Revision 11 M471-10-RH-37. Revision 6 b.

The following characteristics were in accordance with the drawings and Quality Assurance Construction Methods Instruction, QACMI No. M-14, Revision 1, dated March 21, 1978. The welds, located per the drawings, had no visible cracks, lack of fusion, excessive porosity or excessive undercutitng. The lines were properly routed or deviations were controlled by 92i lb9

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as-built (red lined) drawings.

The line components were specified by the drawings. The supports (hangers, etc.)

were incomplete, but controlled. The lines were not identified, however, the licensee indicated that, as construction progresses the lines will be identified in accordance with specification and procedures.

No items of noncompliance were identified.

2.

Instrumentation - Review of Quality Records The inspector reviewed installation records of the subject instrument sensing lines.

Installations were being properly documented on the isometric drawings and KEI QA inspection documented as indicated by the QA inspector's stamp. Liquid penetrant test results on all welds.

No items of noncompliance were identified.

3.

Independent Inspection The inspector reviewed several nonconformance reports to determine current status, are they legible, reviewed by QC personnel and are readily retrievable. Reports adequately describe the noncon-formance. The inspector noted that the disposition instructions were not specific on some of the nonconformance. The licensee stated that they would review this nonconformance procedure for clarity and assure that their personnel understand and follow the procedure. This item is unresolved and will be reviewed during a subsequent inspection.

(358/79-09-01)

Unresolved Matters Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable itens, items of noncompliance or deviations. One unresolved item disclosed during the inspection is discussed in Section II, Paragraph 3.

Exit Interview The inspectors met with licensee representatives (denoted in the Persons Contacted paragraph) at the conclusion of the inspection. The results of the inspection were summarized with one unresolved item being identified. Q ') ?;

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UNITED STATES OF AMERICA 1

NUCLEAR REGULATORY COMMISSION a

BEFORE THE ATOMIC SAFETY AND LICENSING BOARD in the Matter of

)

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Docket No. 50-358 CINCINNATI GAS AND ELECTRIC COMPAh7

)

)

(Wm. H. Zimmer Nuclear Power Plant)

)

DIRECT TESTIM 0hT OF THOMAS VANDEL REGARDING THE PRESSURE TESTING OF DOORS State of Illinois )

)

ss.

County of DuPage

)

Thomas Vandel having first been duly sworn and having provided educational and professional qualifications in written testimony regarding contention number 14, hereby states the following:

At the pre-hearing conference held in the captioned proceeding held on May 21, 22, 23, 1979, an affidavit of Robert Anderson was submitted to the licensing board. A copy of that affidavit is attached.

Mr. Anderson alleges that water tight doors and door frames installed at Zimmer failed pressure testing. Five doors located in the pump house and fire doors located in the reactor building beneath the reactor core all failed to hold a 20 psi pressure when tested because the door frame e= bedded in concrete placement leaked. The licensing board has requested the NRC staff to address this allegation.

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. t During a routine inspection at the Zir=er site en May 22-24, 1979', the Region III office of Inspection and Enforcement inspector determined the following regarding water tight doors.

~

II Mr. Robert Anderson was or, rite during February and March 1978 employed by R. V. Harty Company to instell water tight doors.

2.

The 20 psi pressure test applied by Mr. Anderson was only a contractor proof test and not a licensee quality acceptance test.

3.

The ten doors, as stated, failed to hold the 20 psi pressure due to leaking between the steel angle door frame and the concrete placement f

in which they were embedded.

4.

Subsequent to the time the R. V. Harty Company departed from the site the leaking door frames were successfully repaired utilizing an epoxy plastic material.

5.

Currently the quality acceptance tests are being performed with two frame tests, one at 10 psi and a second at 17 psi being performed, along with a door test at 82 psi (door seal test) also being performed.

6.

The inspector reviewed some test results of tests that had been successfully completed. All door testing had not been completed at the time of the inspection.

Based on the inspector's review of the matter we conclude that the door proble=

is being adequately corrected and we will continue to review this matter.

071cm s

Thomas Vandel Subscribed and sworn to before h/ / day of June 1979.

d me this fW 0 N Sotar lic #

Hy asion expires: /-[-[O

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UNITED STATES OF AMERICA b'UCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY A','D LICENSING BORAD In the Matter of

)

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Docket No. 50-358 CINCINNATI CAS & ELECTRIC COHPANY

)

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(Wm. H. Zimmer Nuclear Power Plant)

)

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DIRECT TESTIMONY OF FEDERICO A. MAURA REGARDING CONTENTIONS No. 15 and 16, CONTROL RODS THICKNESS AND SEALS State of Illinois

)

)

ss.

County of DuPage

)

Federico A. Maura, having first been duly sworn hereby states as follows:

I am employed as a Reactor Inspector in the Reactor Operations and Nuclear Support Branch, Region III, Office of Inspection and Enforcement, Nuclear Regulatory Commission, Glen Ellyn, Illinois. My educational and professional qualifications are set forth immediately below:

Education B.S. Electrical Engineering, Virginia Military Institute 1956.

I have received two certificates from the General Electric Company covering Fundamentals of BWR Operation and BWR Technology, plus one certificate of qualification from the USNRC regarding BWR Advanced Technology Course.

In addition I have held 0)I lbJ E&w 9'

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Operator and Senior Operator Licenses issued by the Atomic Energy s

Commission for the operation of BWR's designed and built by the Allis-Chalmers Company.

Experience I joined the USNRC in November, 1970 as a Reactor Inspector.

In this capacity I have performed inspections of power reactors during the preoperational and startup testing, and operational phases to ascertain conformity with design and other criteria; observed and evaluated the adequacy of licensees' controls and provisions for overall operational safety; management's organizational control, pro;edures and practices, and their relation to the safety of operations; and the status of com-pliance of licensees with licensee provisions, rules, orders, and regulations of the Commission.

I have assisted in specialized inspections of BWR's during the construction phase because of my knowledge on the specific subject.

Prior to joining the Commission I worked eleven years for the Allis-Chalmers Company, Nuclear Energy Division in the design, testing, and startup of their BWR's.

I held the position of Site Manager for the La Crosse Boiling Water Reactor Project.

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From 1957 through 1959 I wtrked for the Duquesue Light Company at the Shippingport Atomic Power Station as Test Engineer during the preopera-

~~

tional testing and startup of that facility and later as Shift Reactor Engineer during the operation of the facility.

The Miami Valley Power Project has raised the following contention:

Contention 15: Control Rods Control rods which must be easily inserted into and removed from the reactor core have been inadequately manufactured so that they do not meet the size specifications for such control rods.

Prior to installation in the reactor vessel the control rods (See Attachment A) are inspected to ensure no damage occurred during ship-ment and/or handling at the site. This site inspection was performed during the period July through October, 1978 by Reactor Controls, Incorporated and consisted of several visual observations as well as measurements as outlined by the checklist used.

(See Attachment B).

The use of the two envelope gauges and the determination of whether to accept or reject a control rod was controlled by General Electric Company, Document No. 22A4387, Revision 4, " Control Rod Handling and Inspection" 07/

165

_4-(See Attachment C).

The 0.280-inch envelope gauge is used to determine if the control rod blade thickness, at any one point in the length of the blade, exceeds 0.280-inches.

The gauge looks like a tuning fork approximately 1-inch wide.

The 0.320-inch envelope gauge is similar to the 0.280-inch gauge except it is approximately 1-foot long and is used to determine if bowir.g exists over a wider area.

During the initial site inspection, conducted by Reactor Controls, Incorporated, of the 137 controls rods 86 did not pass the 0.280-inch thickness envelope gauge.

Of those 86 that did not pass, 4 also did not pass the 0.320-inch gauge used to locate undesirable bowing and the four were rejected.

In accordance with the GE Inspection Procedure (22A4387) a 40-pound force clacp was placed against the blade sheath, adjacent to the high area of the remaining 82 control rods.

The purpose of the clamp was to determine if the local sheath bulge was flexible, and to ensure the absence of foreign matter between the sheath and the poison rods which form the blade. According to the design engineer the use of 40-pounds was somewhat arbitrary since his main interest was determining the flexibility of the bulge.

Seven control rods Jia not pass the 0.280-inch gauge with the 40 pound force clamp applied, one of which was rejected and not placed in the reactor.

The remaining 6 control rods were accepted by the licensee after the General Electric Inspection Procedure was cla:Afied to indicate that the clamp could be placed over

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the high point in question, and the surface area of interest on a con-trol blade was redefined. The clarification was done through the issuance of Field Deviation Disposition Request No. KN-1-286 (See Attachment D),

Because the Reactor Controls, Incorporated and the General Electric Company (FDDR No. KN-1-286) records were not clear as to how the final results (i.e., acceptance of the 6 control rods) were obt31ned, the NRC inspector requested a reinspection of those 6 control rods.

The 6 control rods were reinspected April 10 1979.

The reinspection was witnessed by NRC inspectors. The worse case found was for control rod Serial No. ASIO which on blade No. 2 has a high spot near the edge of the blade which is approximately 0.300-inch in thickness. With the clamp placed adjacent to the high area the thickness decreased to approxi-mately 0.285-inch, and with the clacp over the high spot the thickness was reduced to 0.280-inch or less.

(Refer to Attachment D for location of high spot).

For comparison it should be noted that each control rod has an upper guide roller on each blade (See Attachment A) with a thick-ness of 0.333-inch.

Attachment E shows a typical core cell. According to tne L5. H. Zimmer FSAR the gap between two fuel assemblies in a cell is 0.522-inch which translates into a gap of approximately 0.120-inch between the fuel assembly and a control rod of 0.280-inch thickness.

Control Rod No. ASIO reduces this gap, in the small high spot area, by E

)!

4 less than 10%. On the same date, a meeting was held with the manufacturer's design engineer to discuss the use of 40-pound force clamp and determine the autual control rod blade thickness acceptance criteria limits.

During the meeting, the inspector reviewed a test report containing the results of control rod qualification tests performed by the manufacturer for the purpose of determining the degree of misalignment, channel de formation (wster gap reduction), etc., at which operational performance of the blades would be af fected. During the qualification tests, the control rods were cycled (scrammed, withdrawn, and inst rted) for the expected 20-year design life of the blades, and the wear of the blade sheath and fuel channel was measured.

Based on our review of the inapection records, the reinspection of 6 control rods, the results of the General Electric Company control rod qualification tests, the discussions held with the manufacturer's design engineer, and the knowledge that the systems will be preoperationally tested prior to fuel loading, it is our conclusion that the control rod blades presently supplied to the Zimmer Station are satisfactory because:

1.

The use of 40-pound force clamp was mainly to determine the flexibility of the blade sheath's high area and ensure the absence of foreign matter between the sheath and the poison rods which form the blade.

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The0.280-inchgaugewasacontrolplacedbythedesiknengineerto ensure he was consulted and could analyze any deviations from the desired design parameter.

3.

The blade thickness acceptance criteria could be increased above the maximum thickness measured at the Zimmer Station before the first '. )erational difficulty would be experienced.

4 Preoperational testing of the control rod drive system will demonstrate whether any problems exist which affect the operational performance of the c3ntrol rods.

The Miami Valley Power Project has also raised the following contention:

Contention 16: Control Rod Seals Almost all of the seals on the control rods, which when properly set prevent radioactive water from leaking out when the reactor is shut down for maintenance, do not meet minimum specifications for smooth-ness.

Rough seals cannot set properly, making servicing more difficult and unnecessarily endangering workers and tfa teneral public by causing s

leakage of radioactive water.

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The seal area (See Attachment F) is a machined surface on ilbe control rod bottom casting velocity limiter. When a control rod drive unit must be removed for maintenance the control rod is fully withdrawn at which time the seal area on the control rod seats in its guide tube, similar to the sealing done in ones bath tub.

With regard to the condition of the seal surface on the control rods, the licensee initially rejected one cort.rol rod because of nicks on its seal surface, and it was not placed in the reactor.

The NRC inspectors looked at the seals on tue 6 control blades inspected on April 10, 1979, and found them satisfactory.

Based on the inspector's review of the licensee's inspection records, it is our conclusion that the seals on the 137 installed control rods are satisfactory.

As stated earlier the purpose of the seal is to permit the removal of a control rod drive for maintenance while the reactor is shutdown and depres-surized without having to removed all the fuel and drain the reactor vessel.

Since these are not perfect metal-to-metal seals licensees must be prepared to expect a small amount of leakage until the drive is removed at which time a blind flange can be installed on the control rod drive housing if needed. This small leakage may create an inconvenience to maintenance

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personnel during removal and subsequent reinstallation of control rod drives but, in no case, does it create a safety probleci.

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Attachments: Attachments A through F l 4)

// Aust Federico A. Maura Subscribed and sworn to before f

me this_g g,.aj day of June 1979.

AXSu NY if u Notary Public My Commission expires: /-f-78 h5\\

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ATTACHMENT A

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LOWER CulDE OLLER

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Reactor Controls, Inc.

C0ir" nob ROD III'iPT'?IO11 C]!TKLIST C

SEIAf #

1.

V5IFY THAT ALL ROLL"3S RUU FREE.

2.

VTIFY THAT LATCH 11UT IS TACKED.

3. - VERIFY THAT LATCH IS FREF TO TRAVEL.875" MIllIIIU!I.

h. ' VEIFY DIS?AliC ' FR0!! TOP OF ROD TO BOTToll

. 0F LATCli IS 2.629" I:I!!IIIUM.

5 VEIFY '-'IIAT DISTAUC ' FROM 30TTOM OF COUTROL ROD TO DOTTO:I 0F SOCKT IS.803" MINIIIUM.

6.

211SPC' SEAL SURFACES FOR NICKS, SCRATCHES, FCT.

7 VACCUII FI;TIR ' LZiGTH OF 3LADT.

8.

V21FY ?.!VZLOP2 GAG 2S PASS TIGOUGH THTIR RESPICTIVE LZ:GTHS.

CHICK APPRO?RI!.TI LIHF IF ACCEPTABLE.

BLAD" U.

.23G" GAGF 320 GAG 3 2

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. ESID ' LIHF I;DICATTS 'F.AT CLAI'PIi!3 UAS NXSSSARY III ORDTR TO I' ASS TE' 21!'.~' LOPE Gl.Gr CHTCKS.

C 0!F.il'S :

INSPETID D'/:

DATE:

1. 1 m NOTE: MR13mIllG OF BLAD?S WILL BE PS SK??CH.

C0liTROL ROD S3IAL #

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CONTRCL ROD HANDLE

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ATTACHMENT C EIS IDENT: CONTROL R0D, HDLG & INSP REVISION STATUS SHEET GENER AL h ELECTRIC '

NUCLEAR ENERGY DIVl5tOH MPL - Bil/B13-3070 O seEciricxTioN oRx*:No orsER TvPt INSTattaTI0s spEcirica in, DOCUVENT TITLE CONTROL ROD. HANDLINC, AND INSPECTION LE GEN D-REVISIONS ARE IDENTIFIED BY A SPADE (), )

R E Yl510N5 lC tmm 0

DMF-817 1

Per ECN NE 55528.

Sheets affected:

2 and 3.

Revision identified by a scade (Q).

%%iet 1/9/7s' APR 101975 1mk 2

Per ECN NE 38663.

Sheets affected:

2 and 3.

evisions identified with ApoggyrD FCR a s pa de ( 4 ).

$<CW" JUL g 1975

v. : t H. zi:..'.a P20;E 1 f&A41arx x Ut31 #1 ljp 3

Per ECN NE71043. Sheets 3 and 4 affected. Revisions identified with a spade (4 g G g g N

CHKD BY:

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ECH NE99403 I

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F5s DESCR!PTICN OF GROUPS PRINTS TO

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G E N E R A L h E LE CTRL.C 22A4387 S H. N o.

2 NUCLEAR ENERGY DIVis10N 4

1.

SCOPE 1.1 This document specifies the minimum basic installation requirements for a Cqhtrol Rod into a Reactor Assembly.

$ 2.

APPLICABLE DOCUMENTS 2.1 General Electric Documents. The following listed documents form a part of this specification to the extent specified herein.

2.1.1 Suoporting Documents a.

Control Rod Handling Method 767E711 b.

0.780 Inch Envelope Gase 131C9192 G1 c.

0.320 Inch / Foot Envelope Gage 131C9193 d.

Inspection Clamp 16638544 e.

Del eted f.

Del eted g.

Deleted h.

Deleted 2.1. 2 Supolemental Documents - None 2.2 Codes and Standards None 3.

DESCRIPTION 3.1 Instructions specified herein delineate the inspection and handling require-ments which are necessary to ensure that the Control Rod will be properly handled and installed into the reactor. These requirements form a basis for the CR ins tallation section of the Reactor Assembly Specification and the Reactor Asserely for a particular project.

4.

REQUIREMENTS 4.1 General 4.1.1 The Control Rod (CR) shall be handled in a manner that will protect it from damage during handling into the reactor vessel.

I;

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ATTACINENT C GENER AL h ELECTR1C 22A4387 ss. uo. 3 NUCLEAR ENERGY DIVISION 4

no v.

4.2 Inspection

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4 4.2.1 Inspect 0.280 inch blade thickness by passing the 0.280 inch envelope gage-(131C9192Gl) over the area indicated in Figure 2.

The 0.280 inch thick-ness is a point requirement. Therefore it is acceptable to use a deep thrtat micrometer to inspect any points that do not pass the gage.

D 4.2.2 Del eted.

A, 4.2.3 Inspect 0.320 inch blade envelope for the full length of the sheath by passing the 0.320 inch envelope gage (131C9193) over the area indicated in Figure 3.

4.2.4 If gages do not pass over the sheath, due to local sheath bulge, then use inspection clamp (166B8544) adjacent to inspection gages.

If gage still does not pass over sheath then the control rod is rejectable.

4.3 Handling 4. 3.1 Handling equipment and its use is described in Drawing 767E711. This equipment is used to take the control rod out of the shipping container and prepare it for installation.

4.4 Data 4.4.1 After inspection is complete fill out data sheet (Figure 1) and return to responsible design engineer through the project engineer.

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GENERAL @ ELECTRIC 22r.4387 38 ~o.

s NUCLEAR ENERGY DIVISION REV 4 Final s)

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FIELD DEVIAfl0N DISPOSITION REQUEST issueoeygo

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. -mm NUCLEAR ENERGY DIVISIONS DATE/

FDDR NO.

KN-1-286 1

Zimmer I SHEET OF-m ECT

) DATE ORIGINATED 9/5/78 TELECON OR TWX APPROVAL SY:

P.

Pentz 9/5/78 w.s o z.

s EQUIPMENT IDESCRIPTION AND/OR MPL)

B13-D009: Control Rod DWG. 5PE C. ETC. NO.

SHEET NO.

I REV NO.

DWG. 5PE C. ETC. TITLE 814E935 4

Outline: Control Rod DEscRIFTION OF DEVIATION:

The control rods were inspected as outlined by 22A-4387, R4 with the following changes:

CATEGORYI C

-4.2.1 Reduce the 0.280 inspection a r r.- per Fig A.

CATEGORYli O

-4.2.4 For the following control rods, the inspec-PRIORITY CLAS$1FICATION tion clamp (166B8544) was placed directly EMERGENCY C

on the high point /trea (0.280) and deep URGENT O

M throat micrometer readings (Table 1) as rot / TINE O

outlined by Fig. B was used as the basis for acceptance: S/Ns: A501,.A515,A440,A453, APP OVEDB -

and A420.

OMM 2 7-U DESIGN ENGINEtR DATE A 0.290 high spot was established on control rod, S/N: A510 as detailed in Table 1.

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/of/g SITE IMPACT: Engineering concurrence required uxT 1.S APPLENGidEER prior to WK 39/78.

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SUGGESTED DI5 POSITION 7A 2Whl.- -

1. /4/7f Engineering is to evaluate.the inspection procedurc changes and advise.

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Accept control Rod S/N: A510 as is.

4 1. 9%.

ie, ici n Cost Accounting: GE Responsib.ility.

2~. gTP) f ed 3 -/-79 JEC ANAGER FIELO 6iAN AGER DISAPPROVED BY:

DATE SUPPLEMENT INST REQUIRED YES NO NUMBER FINAL Dl3 POSITION C

@ ECN fA M E' A5 MW E -

I N A l L sq R.EA,$

U W D' OISTRLBUTION CODE St/oN i> tAs fic,pt coa 7kOL R O Q / CHANNEL / WTERFeRG._

INTERNAL EXTERNAL W s L.L.

uo r occ og P210 2 Yo E^/ O o f cMRA/WC.

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D'HER PLANT 3 AvvECTED VES O IDENTIFY

%O Q SAFETY / RELIABILITY REVIEW REQUIRED:

YEs O NoR FIELO WOR ORDER NO.

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