ML20237L298

Rev 0 to Nuclear Engineering & Operating Procedure Neo CS-1, Evaluation of & Reporting of Items/Events Under 10CFR21 & 10CFR50.55(e)

ML20237L298
Person / Time
Site:	Comanche Peak
Issue date:	10/21/1985
From:	Counsil W TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20237K807	List: IR 05000445/1985007 IR 05000445/1985014 IR 05000445/1986008 IR 05000446/1985011 ML20237K804 ML20237L110 ML20237L125 ML20237L136 ML20237L151 ML20237L159 ML20237L179 ML20237L203 ML20237L268 ML20237L272 ML20237L298 ML20237L321 ML20237L367 ML20237L407 ML20237L453 ML20237L464 ML20237L493 ML20237L577 ML20237L797 ML20238A290 ML20238A390 ML20238A485 ML20238A535 ML20238A548 ML20238A573 ML20238A583 ML20238A596 ML20238A609 ML20238A642 ML20238A654 ML20238A667 ML20238A719 ML20238A749 ML20238A782 ML20238A833 ML20238A839 ML20238A868 ML20238A883 ML20238A903 ML20238A923 ML20238A966 ML20238A974 ML20238B015 ML20238B131 ML20238B210 ML20238B251 ... further results
References
NEO-CS-1, NUDOCS 8708200164
Download: ML20237L298 (20)

Text

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JUL 19 'E6 13:01 LICENSING-TUGCO PAGE.02 1

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4 TEXAS UTILITIES GENERATING COMPANY NUCLEARENGIkEERINGANDOPERATIONSPROCEDURE NE0 CS-1 EVALUATION OF AND REPORTING 0F ITEMS / EVENTS UNDER 10CFR21 AND 10CFR50.55(e)

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l APPROYED ctg' MJ, Executive Vice President Nuclear Engineering and Operations REVISION 0 - Effective 11-1-85 DATE

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CONCURRENCE b[M/d'[(

D1 Mctor,~ Quality Assurance

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JUL 19E5>13:01 LICENSING-TUGC0 PAGE.03 s.

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TEXAS UTILITIES GENERATING COMPANY NUCLEAR ENGINEERING AND DPERATIONS PROCEDURE NEO CS-1 o

EVALUATION OF AND SEPORTING 0F ITENS/ EVENTS UNDER 10CFR21AND10CFR50.55(e) 1.0 PURPOSE The purpose of this procedure is to delineate the method of determining the deportability and subsequent reporting of significant items / events to the Nuclear Regulatory Commission (NRC) as required by 10CFR21 and 10CFR50Section50.55(e).

2.0 APPLICABILITY 2.1 This procedure is applicable to all Texas Utilities Generating Company employees during desigs, construction. and operations activities of-Comanche Peak Steam Electric Station (CPSES).

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2.2 This procedure is applicable to systees and structures identified in the Final Safety Analysis Report Arpendix 17-A which are Safety Class 1, 2, or 3; Class IE; or seismic Category 1.

It also applies to services which could create a substantial safety hazard if they were not i

adecuately performed.

2.3 This procedure applies to items and activities identified which are required to be evaluated for deportability to the NRC except those reportable under 10CFR50.72 and 10CFR50.73.

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3.0 REFERENCES

3.1 10CFR21 3.2 10CFR50.55(e)

I 4.0 DEFINITIONS 5

4.1 Potentially Repertable Item / Event l

t An item / event which is of suih sigriificance that a complete analysis may find it reportable.

NEO CS-1 Rev. 0 Datel 11/1/85 Pace 1 of 8

JUL 19 '8E.13:02 LICENSING-1UGCO PAGE.04 4.2 Reportable Item A significant item / event which has been evaluated pursuant to this procedure and the applicable regulations 'an( is determined to be reportable to the Nuclear Regulatory Comeission!

4.3

$1onificant item / Event An item / event whit.h has a significant adverse effect on the safe operation of the facility.

5.0 RESPONSIBILITIES 5.1 All TUGC0 Emoloyees Responsible for informing the Site Coordinator of all potentially significant items / events as they are identified.

5.2 Executive Vice President, Nuclear Engineering and coerations (NED)

Responsible for the approval of this procedure and subsequent revisions and for the final review, approval, and issuance of all written reports.

5.3 Director. Quality Assurance (OA)

Responsible for final concurrence with this procedure and subsequent l

revisions, q

l 5.4 Licensing Coordinator Responsible for trackin5 and verbal notification of the NRC.

5.5 Site coordinator (the Construction Staff Engineer)

Responsible for doc eenting potentially significant items / events.

5.6 Vice President. Nuclear Ooerations Responsible for ensuring that his organization has procedures in effect, as necessary, that will implement the action prescribed in Section 6.0, 5.7 Vice President, Engineering and Construction and CPSES Project General Manager Responsible for ensuring that his organization has procedures in effect, as necessary, that will implement the action prescribed in Section 6.0.

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l S8 Vice President (responsible for Quality Assurance, Licensino and Nuclear l

Fuels)

Responsible for ensuring that his organization'hati procedures in effect, as necessary, that will implement the action prestribed in Section 6.0.

l 5.9 Manager. Perchasing and Stores Y-Responsible for seeing that procedures exist within his organization which ensure that purchasing documents issued meet the requirements set forth in Section 6.0.

In addition, he will ensure that his personnel involved with the pt.rchast of materials for CPSES are aware of their responsibilities as defined by this procedure.

6.0 INSTRUCTIONS 6.1 initiation Anyone detecting a potentially significant item / event nr being notified by a vendor of a.potentially reportable or reported Part 21 deviation shall notify the Site Coordinator either verbally or in writing. The Site Coordinator may request that the infomation be provided in writing.

6.2 Documentation 6.2.1 Upon receipt of information concernirg a potentially sigMfi-cant item / event, the Site Coordinator will document the problem on the form shown in Figure 7.2.

6.2.2 The Site Coordinator shall maintain a log which allows him to ascertain when all necessary records have been assembled.

6.3 Distribution for Evaluation /Reoort 6.3.1 The Site Coordinator shall transmit the documented information of Section 6,2.1 to the appropriate mana This manager may be (but is not limited to)ger for evaluation.

• Manager,-QA

• Manager, Quality Control (QC)

• Manager Nuclear Operations

• Manager, Plant Operations

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• Manager Engineering

• Assistant Managers. Engineering If the assigned manager and the Site Coordinator disagree on the assignment, the matter shall be resolved by the Manager, j

Licensing.

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6.3.2 The Site Coordinator shall distribute copies of the Po'tentially Reportableitem/Eventform(Figure 7.2)to a) Vice president (responsible for QA, Licensing, and Nuclear l

Fuels) l b?VicePresident,NuclearOperations cs Vice President Engineering and Construction and CPSES Project General Kanager, ' ':,'

sl Director,QA d

Licensing Coordinator 6.3.3 The $tte Coordinator shall maintain a file copy for tracking purposes.

6.4 Evaluation 6.4.1 The evaluator shall make an evaluation of the deportability of the potentially reportable item / event.

6.4.1.1 Reference should be made to the log (ic contained in

" Flow Chart - 10CFR21 Deportability" Figure. 7.3) and

" Flow Chart - 10CFR50.55(e) Deportability" (Figure 7.4).

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6.4.1.2 The evaluation shall be documented on the form shown l

in Figure 7.2.

6.4.2 If the evaluator recommends that the potentially reportable item / event is to be reported, a written report shall be-prepared.

6.4.2.1 The report shall contain all information required by 10CFR21.21 even though the report may concern an item / event reportable under 10CFR50.55(e).

6.4.2.2 The report shall contain the information shown in Figure 7.5.

6.4.3 Upon completion of the evaluation, all documentation shall be transmitted to the appropriate individual below:

1)

The Vice President Er.gineering and Construction and CPSES Project General

Manager, if the evaluator was, a construction manager.

2)

The Vice President, Nucteer Operations, if' the evaluator was an operations manager.

L NEO C5 1 Rev. 0 Date: 11/1/85 Pann a of A

3) ~ The 'Vice President (responsible for Quality Assurance, Licensing and Nuclear Fuels), if the evaluator was a Manager of QA, Licensing, or Nuclear Fuels.

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6.5 Determination of ReDortability 6.5.1 The individuals identified. Jn Section 6.4.3 shall document their agreement or disagreement with the evaluation results on the form shown in Figure 7.2 and return the documentation package to the Site Coordinator.

6.5.2 The Site Coordinator shall log the receipt of the evaluation and transmit the package to the Licensing Coordinator.

6.5.3 If the evaluation recommends that the item / event be. reported, the Licensing Coordinator shall transmit the package to the Executive Vice President, NE0 for a final decision.

6.5.4 If the evaluation is that the item / event is not reportable, the Licensing Coordinator shall transmit the package to. the Vice President for concurrence or Executive Vice President, NEO.

6.5.4.1 If the Executive Vice President or the Vice President (responsible for QA, Licensing, and Nuclear Fuels) agrees with

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the nonreportable status, his decision shall be documented on the form shown in Figure 7.2 and the Licensing Coordinator shall:

1) Notify the NRC (if it had previously been notified j

that the item / event was potentially reportable)

2) Log any applicable information.

3) Transmit the package back to the Site Coordinator 4

for record retention purposes.

6.5.4.2 If the Vice President (responsible for QA, Licensing, and Nuclear Fuels) disagrees with the nonreportable status (he thinks the item / event is reportable), his decision shall be documented on the form shown in Figure 7.2 and the t.icensing Coordinator shall transmit the package to the Executive Vice President, NE0 for a final decision.

NE0 C5-1 Rev. 0 Date: r.,1/85 11/

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6.6 Tracking for Evtluation 6.6.1 The Licensing Coordinator shall track the documentation packages for a 30 day completion.

6.6.2 If, at the end of the 30-day period the package has not been completed, he shall contact.the appr,opriate party to obtain an estimated completion date.-

6.6.2.1 He shall update his log.

6.6.2.2 He shall notify the NRC according to Section 6.7.

6.7 Noti fication 6.7.1 The Licensing Coordinator shall notify the NRC and sh'all report the.' item / event as potentially reportable along with the esti.

mated completion date of the evaluation. He shall document the notification.

6.7.2 Upon receipt of a package which the Executive Vice President.

NED has decided is reportable, the Licensing Coordinator shall notify the NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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6.7.1.1 He shall report the item / event as reportable.

6.7.1.2 He shall document the notification.

6.7.1.3 Ne shall 1cg any applicable information.

6.8 Written Repor_t 6.8.1 The Licensing Coordinator shall be responsible for the submittal of any written reports.

6.8.1.1 The report shall contain all information required by 10CFR21.21 even though the report. may concern an item / event reportable under 10CFR50.55(e).

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6.8.1.2 The report shall contain the information shown in Figure 7.5.

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j 6.8.2 The transmittal letter for the written report shall be signed by the Executive Vice President, NEO.

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1 6.9 Report Distribution Copies'of the report shall be transmitted to the NRC in accordance with the requirements of 10CFR21 or 10CFR50.55(e) as appropriate.

Internal distribution will be made at the direction of the Manager, Licensing.

6.10' Records 6.10.1 One copy of the report and any additional documentation that has been generated during corporate review shall be distributed to the Site Coordinator for record retention.

6.10.2 The' Site Coordinator shall maintain recon!S indicating the disposition of any items / events, that have been beught to his attention as a result of this procedure. These records shall be classified as quality records and shall be maintained for

.the life of the plant.

6.11 Letter to Board of Directors The Licensing Coordinator shall assure that the letter from the Chairman of the Board and Chief Executive Officer. Texas Utilities Electric Company (TUEC). and Chief Executive to the Board of Directors (shown in Figure 7.6) is retransmitted to the Board during the first quarter of

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each year.

6.12 Postino Each individual Vice President of activities within the scope of this procedure shell assure the following documents are posted in a conspicuous position:

I a A copy of this procedure b 10CFR21 c Section 206 of the Energy Reorganization Act of 1974 6.13 Procurement Documents Procurement documents issued which pertain to safety-related equipment or services for design, construction, or operation and which are under jurisdiction of 10CFR21 (i.e., not coussercial grade) shall contain the following provisions:

a) The provisions of 10CFR21 apply.

b) A copy of any report furnished to the NRC in compliance.with the

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provision of 10CFR21 concerning equipment or services intended for use in any Texas Utilities Company System nucl. ear power plant must be

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transmitted to the Executive Vice President, NEO at the same time and in the same manner in which the report is furnished to the k

Connission.

NE0,CS-1 Rev. O Date:n...,11/1/95 so

7.0 FIGURES Figure 7.1 Procedure Flowchart Figure 7.2 Potentially Reportable Item / Event Form Figure 7.3 Flow Chart 10CFR21 Deportability Figure 7.4 Flow Chart 10CFR50.55(e) Repor,tability Figure 7.5 Reporting Requirements

Figure 7.6 Latter to the Board)f Directors 8.0 ATTACHMENTS None NE0 CS-1 Rev. 0 Date: 11/1/85 Dana R af 8

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FIGURE 7.2 j

POTENTIALLY REPORTABLE ITEM / EVENT

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SOURCE OF INFORMATION STATEMENT OF' PROBLEM l

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EVALUATION ASSIGNED TO EVALUATION DUE DATE:

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EVALUATION:

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RECOMMENDATION:

REPORTABLE OYES C NO / APPLICABLE TO:

O 50.55(e) w O PART 21 O BOTH BY:

DATE:

CONCLUSION:

REPORTABLE O YES O NO

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CONCLUSION:

REPORTABLE O YES C NO / APPLICABLE TO:

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Figure 7.3 I

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Flow Chart 10CFR21 Deportability Deficiency [dentified,

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Deficiency in Plant Security which, on the basis of an evaluation.

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could create a substantial safety i

hazard to Failure to comply with Atomic Energy Act of 1954, as amended, or any applicable rule, regulation, order Yes --

or license of the NRC which, on the.

basis of an evaluation, could create a substantial safety hazard to No Defect in a lasic Component (as Evaluation Not Complete Within--

t 30 Days V

Item Not Report to NRC Within Repor'able 24 4ours 5-Day Fornal Report Rev. 0

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Date:11-1-85 Pagt 7.3 - 1 of 1 NE0 CS-1 i

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Figure 7.4

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10CFR$0.55(e) Deportability Deficienc,y Identified No Had it remained uncorrected, could have adversely affe ed the safety of operatjons' Yes i

Deficiency Represents a Significant

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Breakdown in the QA Program No Deficiency is a Significant Deficiency in final design as approved and released for Construction such that the Yes-design does not conform to the criteria and bases in SAR or Construction Permit No Deficiency is a Significant Deficiency in construction.

damage, or deviation from performance specifications which No-will require extensive evaluation, repair, or redesign to meet criteria and basis in SAR or Construction Permit or to restore the item to meet its intended Safety Function '

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Yes Evaluation Not Complete within -

30 Days V

) tem Not Report to NRC Within Reportable 24 Hours 30 Day Fomal Report l

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Date: 11-1-85 l

Page 7.4 - 1 of 1 1

NEO CS-1 i

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Figure 7.5 Reporting Requirements 1.

Name and address of the individual or individuals informing the Comission.

2.

Identification of the facility, the activity or the basic components supplied for such for:ility or such activity within the United States which fails to comply or contains a. defect.

3.

Identification of the fim constructing the f acility or supplying the basic component which fails to comply or contains a defect.

4.

Nature of the defect or failure to comply and the safety hazard which is created or could be created by such defect or failure to comply.

5.

The date on which the information of such defect or failure to comply was obtained.

6.

In the case of a basic component which contains a defect or fails to comply, the number and -

location of all such components in use at, supplied for or being supplied for one or more

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facilities or activities subject to the regulations in 10CFR21.

7.

The corrective action which has been, is being or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action.

8.

Any advice related to the defect or failure to comply about the f acility, activity or basic component that has been, is being or will be given to purchaser or, licensees.

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Rev. O Date: 11-1-85 Page 7.5 1 of 1 NEO CS-1

Figure 7.6 l

TEXAS UTILITIES GENERATING COMPANY

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Members TUEC Board of Directors Mr. M.D. Spence TUEC 10CFR21 Policy Texas Utilities Electric Company (TUEC) recognizes and accepts the responsibility of compliance with Title 10 of the Code of Federal Regulations. Part 21.

This regulation applies to any individual Director or responsible. officer of TUEC which is the owner, constructor and operator of Comanche Peak Steam Electric Station (CPSES).

The responsible officer of TUEC is Mr. M.D.

Spence TUGC0 Division i

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President.

Mr. Spence, at my direction, has established a program for TUEC that implements the, requirements of Part 21.

In the event that TUEC Directors become aware of a potential defect or deviation that could create a safety hazard at the CPSES Facility, it is requested that they bring the issue to the attention of Mr. M.D. Spence. Your assistance in this matter l

will help assure the safe and reliable operation of CPSES.

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TT.XAS UTILITIES GENERATING COMPANY OFFICE MEMORANDUM Te_ _

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Din Attached for ycur informatQn, J& ease [lfAd the following proced ures:

WE0 3.01 Corrective Action WE0 3.02 Conditional Release grE0 3.05 Reporting and Control of Nonconforusnees REO 3.06 Reporting and Control of Deficiencies These recently approrved procedures represent a significant chan existing programs describing the control of non-compliant items or ge to conditions.

The effective date of these procedures. ninety (90) days subsequent to issue, has been established to allow an orderly tran through accomp1.ishing the follwing activities:

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If required in the applicable NE0 procedure, all existing Construction NCRs, Startup Test Deficiency Reports, TUCC0 D Deficiency Reports, and Operations NCRs and Deficiency Rep

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may be converted to NCRs or Deficiency Reports (DRs) in the e an approved disposition has not been issued.

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All monconformance/ deficiency documents shall be converted t WCRs or DRs, as describad in the NE0 procedures, no later than o

six (6) months af ter the effective date.

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within the organtastional structure of the Execu and reviesd as necessary to incorporate spec e

4 Key management / supervisory personnel, inclu procedures.

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comm1teents to the Inte shall be reviewed ta,113ht of des toeved is acgordance with the-DEO proc r

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Proevrement documents for contractors working on site shall be supplemented as necessary to require the implementation of these procedures to order to estabitsh a single, manconformance/ deficiency control program. project-wide Your prompt attention to these activities is essential.

controlled distribution of these procedures will be acceeplished asPlease note, specified la the NE0 Polfetes and Procedures Manual.

54du W. C. Counsil WGC:mac Attachments c - w/o attachments:

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Me rritt E. L. Scheppele R. R. Westf ahl R. C. Iotti R. L. Crabb W. R. Deatherage R. D. Centry J. D. Edwards R. E. Kahler B. J. Cheatham B. C. Sciudit J. F. Streeter C... reeles REGElYE6l T. C. Tyler R. A. Jones T. L. Cosdin g[p g319S$

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March 6, 1985 TXX-4409 Mr. 0.R. Hunter, Chief Reactor Project Branch 2 U.S. Nuclear Regulatory Comission l

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Office of Inspection & Enforcement 611 Ryan Plaza Drive, Sulte 2000 9-Docket hos.: 50-445 Arlington, TX 76012 50-446 COMANCHE PEAK STEAM ELECTRIC STATION VENTILATION EXHAUST DAMPERS QA FILE: CP-84-27, 50AR-152 FILE NO.: 10110

Dear Mr. Hunter:

In accordance with 10CFR50.55(e), we are submitting the enclosed report of actions taken to correct a deficiency regarding ventilation exhaust dampers that have been observed to be designed / installed to fall closed which would divert radioactive releases to rooms in which operator action is required. We have submitted interim reports logged TXX-4333, TXX-4365 and TXX-4379, dated October 12, 1984, November 20, 1984 and December 17, 1984, respectively.

Supporting documentation is available at the CPSES site for your inspector's review.

Very truly yours, Billy R. Cle/A Al DL ments BRC:tig cc: NRC Region IV - (0 + 1 copy's Director, Inspection & Enforcement (15 copies)

U.S. Nuclear Regulatory Comission Washington, DC 20555

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ATTActetEhi VENTILA110N EXHAUST DAMPERS Description During disposition of a design deficiency report by site engineering, the designed failure mode for several ventilation dampers was observed to be fail-closed. Also, it was determined that when in the post-accident mode of operation, airborne radioactive releases could be diverted from normal exhaust paths to rooms required for access for operator actions.

Evaluation of this issue involved an assessment of,-ttie primary plant ventila-tion scheme. Initial concerns regarding the effects-of radiation were dispelled without affecting offsite releases or operator action. Compromise of the class 1E power supply, also an initia.1 issue, was resolved without adverse affects; however, the review concluded the effects of increased temperature and hydrogen releases within the ventilation system were unacceptable. Specifically, increased temperatures during an accident could esceed qualification parameters and result in indeterminate operability of the electrical equipment in the affected areas.

In addition, radioactive, hydrogen-rich gases could result in an explosive mixture of hydrog.n in the event of an inadvertent leak from the boron recycle holdup tanks.

Safety implications in the event the deficiency had remained undetected, the ability of the operator to safely perform essential functions could be impaired or pro-hibited. In addition, the operability of this equipment required under r

accident conditions could not be assured.

Corrective Actions In order to assure operability of the affected electrical equipment, the ventilation system will be modified to provide adequate heat removal.

Eighty-eight (88) dampers will be removed or modified tar accomplish the system rework. An additional thirty-three (33) dampers will be involved in the effort in order to minimize the impact to scheduling and engineering.

Further activities, such as insulating additional piping, have been implemented to reduce heat generation.

Equipment qualification reports of the affected electrical equipment *ere being reviewed and updated to incorporate a 72-hour loss of ventilation.

The relocation of components and instruments is also underway to ministre 4'

the impact of the qualification program, t.,

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11A.6409 3/6/65 Page 3 For Unit 1, design, construction, and inspection activities for the damper effort are scheduled for completion in July 1985. Equipment qualification report, F5AR, and specification updates will be complete in September l

1985. Unit 2 activities will be completed prior to startup testing.

J In order to mitigate the effect of potential hydrogen releases, explosion-proof lighting will be installed where applicable. A spect,al HVAC test to ensure gas removal will also be performed.

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November 5, 1984-TXX-4350 NOV-624i

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Mr. 0.R. Hunter, Chief Reactor Project Branch 2 Y-U.S. Nuclear Regulatory Conrnission

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p Office of Inspection and Enforcement 611 Ryan Plaza Drive, Suite 1000 Oc"';et No.: 50-445 Arlington, TX 76012 COMANCHE PEAK STEAM ELECTRIC STATION CONTROL ROOM HVAC QA FILE: CP-84-29, SDAR-154 FILE NO.: 10110

Dear Mr. Hunter:

in accordance with 10CFR50.55(e), we are submitting the enclosed report of actions taken to correct a deficiency regarding the safe shutdown analysis evaluations which did not consider a potential fire which could effect habitability of control room and alternate shutdown capability. We have submitted an interim report logged TXX-4335 dated Octcber 12, 1984.

Supporting documentation is available at the CPSES si}e for your inspector's I

i review.

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Very truly yours, bl db i.<N BRC:tig Attachment (c: NRCRegionIV-(0+1 copy)

Director, inspection & Enforcement (15 copies)

U.S. Nuclear Regulatory Consnission Washington, DC 20555

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TXX 43SO November 5, 1984 i

ATTACHMENT l

Control Room HVAC Description t

i Review by site engineering has revealed an unreviewed fire scenario.

The l-ef fect of the unpostulated fire could preclude control room habitability due to environmental conditions. In addition, several equipment functions g

necessary to control the plant from the hoa sh'utdown panel (HSP) could be rendered iroperable by the same fire.

Mr.

Sa,fe_tL implications i

in the event the conditions had remained undetected, the ability of the i

operator, the safety monitor and control plant functions would be adversely j

affected.

i Corrective Action in order to mitigate the consequence; of the fire, all essential sables and i

components will be afforded protection by the installation of one-hour fir?

i barrier material and fire detection and suppression capabilities in accordance with 10CFR50, Appendia R, Section Ill.G.2.c and approved CPSES deviations.

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March 11, 1985 TXX-4437 Mr. D.R. Hunter, Chief

.-f Reactor Project Branch 2 T-U.S. Nuclear Regulatory Conrnission Office of Inspection & Enforcement 611 Ryan Plaza Drive, Suite 1000 Docket Nos.: 50-445 Arlington, TX 76012 50-446 COMANCHE PEAK STEAM ELECTRIC STATION CONTAINMENT #Nou HEADLK

~ TION VALVES QA ILE:

CP-85-0,4, SD 165

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Dear Mr. Hunter:

In accordance with 10CFR50.55(e), we are submitting the enclosed report of actions taken to correct a deficiency regarding containment spray header isolation valves that open in less than 20 seconds.

Supporting documentation is available at the CPSES site for your Inspector's review.

Very truly yours, hhh.

A.R.Clements B

BRC:tig Attachment cc: NRC Region IV - (0 + 1 copy)

Director, Inspection & Enforcement (15 copies)

U.S. Nuclear Regulatory Comission Washington, DC 20555 l

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TkX-443'i 3/11/85 Pege 2 ATTACHMENT CONTAINMENT SPRAY HEADER ISOLATION VALVES

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Description:==

During preparation of proposed plant Technical Specifications, a discrepancy was identified by site engineering regarding the opening time of the containment spary heeder isolation valves.. A subsequent review was perfonned which encompassed FSAR comitments, specification and test data, and review of calculations by the Architect /Er.gineer..Preoperational testing confirmed the opening times of the valves were consistent with the acceptance criteria; and, the acceptance criteria, less than twenty (20) seconds, properly reflected the procurement specifications.

l Review of the calculations, indicated that slow opening of the valves (approximately 117 seconds) was required to prevent starting trips of the containment spray pump motors under design conditions. Tne discrepancy between the installed valves (and specifications) and the pump motor operating parameters results in a conrnon mode failure which could defeat both safety train functions. The concern is applicable to Units 1 & 2.

l Sl ety implications f

In the event the deficiency had remained undetected, operability of containment l

spray system could not be assured.

Corrective Action The operators for the valves and all associated design documents will be modified to slow the opening time, in order to minimize these modifications, the valve logics will also be revised to close the recirculation valves coincioent with the opening of the isolation valves. These actions will l

further ensure protection of pump motor operation.

Based upon procurement schedules for the valve operator replacement phrts, these actions should be completed no later than July 31, 1985 for Unit 1.

Unit 2 activities will be performed prior to startup testing, i

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Mr. D.R. Hunter, Chief T-Reactor Project Branch 2 U.S. Nuclear Regulatory Commission Office of Inspection & Enforcement 611 Ryan Plaza Drive, Suite 1000 Docket Nos.: 50-445 Arlington, TX 76012 50-446 COMANCHE PEAK STEAM ELECTRIC STATION DG ENGINE CONTROL PANEL AIR FILTER BOWLS QA FILE: CP-85-05, SDAR-166 FILE NO.: 10110

Dear Mr. Hunter:

in accordance with 10CFR50.55(e), we are submittdng the enclosed report of actions taken to correct a deficiency regarding the bowl for the air filter in the engine control panel that may be underrated for the pressure seen by the filter under normal operation.

Supporting documentation is available at the CPSES site for your Inspector's

. review.

Very truly yours, RECEIVED Meh FEB 2 01985 Texas Utilities Generating Co.

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Hooton Director, Inspection & Enforcement (15 copies 1calcer

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l ATTACHMENT DG ENGINE CONTROL PANEL AIR FILTER BOWLS Description The supplie-of the diesel generators for CPSES, Delaval, has advised us of a concern identified at another nuclear facility dealing with the air filter bowl in the engine control panel.- Original supplier purchase specifications for the air filter bowl required a polycarbonate transparent bowl rated at 250 psig.

Subsequently, the specification was revised to specify a metal bowl and revised ratings were issued for both the polycarbonate and metal bowls.

The current rating of the polycarbonate bowl (150 psig at 1250F) is not acceptable.

Review by jobsite pesonnel has confirmed the concern is applicable to CPSES.

Considering the installation of the polycarbonate bowls, f ailure could occur before starting and during engine operation due to the loss of air pneumatic centrols.

Safety Implications In the event the defect had remained undetected, operability of the diesel generators could not be assured.

Corrective Act The existing polycarbonate bowls will be replaced with metal bowls in order to assure proper pressure rating of the engine control panel air filter system.

Procurement activities for the replacement parts have been initiated.

Installation activites will be monitored by the master data base items issued for the replacement of the defective items.

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M. D. Spence B. R. Clements D. N. Chapman R. L. Ramsey

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D. Frankum J. W., Beck J. S. Marshall c"

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j F. B. Shants J. B. George M. R. McBay T. L. Smart R. E. Kahler N. S. Reynolds R. T. Jenkins l

Records Center i.

Institute of Nuclear Power Operations 1100 Circle 75 Parkway Suite 1500 Atlanta, GA 30339 i

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TEXAS UTILITIES GENERATING COMPANY mM YW AY TOWEN. 400 SWORTN OLITE STREET. L.B. 88. DALLAm. TEEAs 19801 RECEIVED

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Apri1 10, 1985 APR 151985 TXX-4454 Texas Utilities Generating Co.

CPSES Engineering Div.

Mr. D.R. Hunter, Chief "r.

Reactor Project Branch 2 U.S. Nuclear Regulatory Comission t

Office of Inspection & Enforcement 611 Ryan Plaza Drive, Suite 1000 Docket No.: 50-445 Arlington, TX 76012 COMANCHE PEAK STEAM ELECTRIC STATION INSTRUMENT FITTING LOCATIONS QA FILE: CP-85-11. SDAR-172 FILE NO.: 10110

Dear Mr. Hunter:

In accordance with 10CFR50.55(e), we are submitting the enclosed report of actions taken to correct a deficiency regarding a drawing requirement that no tube fittings be installed between two 450 bends in instrumentation tubing.

Supporting documentation is available at the CPSES site for your Inspector's review.

Very truly yours, 1

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Director Inspection & Enforcement (15 copies) 8*nSe Ruimerman U.S. Nuclear Regulatory Commission Madden Stevens Washington. DC 20555 gg lW m

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TXX-4454 4/10/85 Page 2 ATTACHMENT INSTRUMENT FITTING LOCATIONS Description During a systematic review by(TUGCo Nuclear Engineering to incorporateCMC's) into th Component Modification Cards of limiting conditions have been identified regarding'the instrumentation tubing on CMC 57487.

Specifically, the design change contains a note regarding the location of fittings on tebing installations containing bend radii between supports.

The intent of the note was to state fittings were not allowed between supports in the event the length of tubing between the bend radii exceeds eleven inches (11"). However, the note has been interpreted incorrectly in the construction and inspection proce n to not allow fittings between the bend radii if the length of tubing (between the bend radii) is greater than eT We'n inches (11").

Misinterpretation of the note has resulted in instrumentation installations which may be unacceptable. The. improper placement of fittings could amplify stress intensity factors beyond the current seismic evaluation. As a result, associated instrumentation and control systems could be rendered inoperable due to failure of the tubing and supports during a seismic event. The deficiency has been documented on a site nonconfomance report.

Safety Implications in the event the deficiency had remained undetected, the capacity of the operator to safely monitor plant conditions cou'Id be adversely affected during a seismic event.

Corrective Action Using the instrument seismic support drawings, engineering w_ilj identify i

the specific installations for which specific as-built infomation will be required for evaluation. These installations (approximately 100) will be re-inspected by site quality personnel. For the purposes of identification and inspection, the tubing length requirement between bend radii will be reduced to eight inches (8").

Based upon the results of the inspection, individual installations will be evaluated on a case-by-case basis. Rework, if required, will be specified and implemented. The subject note has been clarified to preclude further l

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recurrence..

In order.to validate the identification and inspection process, engineering personnel will perfom a field verification of an additional 100 installations.

l The sample will be expanded as required.

L Our current schedule indicates engineering and QC activities should be complete in mid-May. Construction activities will be dependent upon release of systems by operations.

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M. D. Spence B. R. Clements J. R. Wells R. L. Ramsey

'J.C.Kuykendal(

h T W rritt R. E. Camp-D.'Frankum J. W. Beck P. E. Halstead J. S. Marshall-J. Cummins F. B. Shants J. B. George L. M. Popplewell-T. L. Smart R. E.'Kahler N. S. Reynolds R.'T. Jenkins Records Center Institute of Nuclear Power Operations 1100 Circle 75 Parkway, Suite 1500 Atlanta, GA 30339 l

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TXX-44bo g'needng Div.

Mr. 0.R. Hunter, Chief Reactor Project Branch 2 U.S. Nuclear Regulatory Commission l

Office of Inspection & Enforcement 611 Ryan Plaza Drive, Suite 1000 Docket Nos.: 50-445 Arlington, TX 76012 50-446 COMANCHE PEAK STEAM-ELECTRIC STATION AUXILIARY UWAlt.K g-RE CONTROL QA FI CP-85d 27 5DAR-173 FILE N0.: 10110

Dear Mr. Hunter:

In accordance with 10CFR50.55(e), we are submitting the enclosed report of actions taken to correct a deficiency regarding a preliminary flow balance calculations that have been identified as unacceptable setpoints for the auxiliary feedwater system.

Supporting documentation is available at the CPSES site for your Inspector's review.

Very truly yours.

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Director, Inspection & Enforcement (15 copies) 8:5RC - M U.S. Nuclear Regulatory Comission LdE-/C[

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TXX-4456 4/15/85 Page 2 ATTACHMENT AUXILIARY FEE 0 WATER PRESSURE CONTROL Description-As a result of an unrelated evaluation, preliminary flow balance calculations

-by TUGCo Nuclear Engineering (TNE) have identified unaccept4ble setpoints involving the auxiliary feedwater control valve u Dur review indicates the condition can be attributed to two (2) issues. Initially,-the setpoint calculations did not properly reflect the certified " pump" curve submitted by the supplier which reflects actual planp operation. In addition, a site-generated system modification has not been considered in the existing calculations.

The auxiliary feedwater system control valees regulate supply flow to the steam generators by sensing the discharge pressure of the auxiliary feedwater l

pumps. The existing setpoints for these valves would result in flow approx-1 imately 17% below that required. These conditions would restrict steam generator supply to below prescribed technical specification levels and would unacceptably decrease decay heat removal for postulated accidents and safe shutdown.

As a result of the status of Unit 2 construction, the concern is limited to Unit 1.

Safe g Implications in ine event the concern had remained undetected, the ability of the operator to achieve safe shutdown or mitigate an accident could be impaired.

Corrective Actions Engineering activities which include hydraulic calculations and instrumentation calculations will result in revised setpoint and scalling documents. These efforts should be concluded by May 31, 1985..

Recalibration of the control valves will be accomplished by Operations. As governed by the technical specifications, these activities will be accomplished j

prior to system operations in Modes 1, 2 or 3.

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CPSES Engineering Div.

TXX-44 7, 1985

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Mr. D.R. Hunter, Chief Reactor Project Branch 2 4-U.S. Nuclear Regulatory Comission Office of Inspection & Enforcement 611 Ryan Plaza Drive, Suite 1000 Docket Nos.: 50-445 Arlington, TX 76012 50-446 COMANCHE PEAK STEAM ELECTRIC STATION UNDETECTABLE FAILURE IN SAFETY FEATURFC ^ CT' f,T" SYSTEM J

QA FIL C rP-85-13 3 AR-173 FILE NO.: 10110

Dear Mr. Hunter:

In accordance with 10CFR50.55(e), we are submitting the enclosed report of actions taken to correct 4 deficiency regarding failures that exist in the l

safety features actutition system.

Supporting documentation is available at the CPSES site for your Inspector's review.

Very truly yours, k(]/Nw BRC:tIg

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cc:

NRC Region IV - (0 + 1 copy) g,/' y Director, Inspection & Enforcement (15 copies)

U.S. Nuclear Regulatory Comission V',"k.-f-/

Washington, DC 20555 MA

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TXX-4457 4/15/85 Page 2 ATTACHMENT UNDETECTABLE FAILURE IN SAFETY I

FEATURES ACTUATION SYSTEM Description The NSSS supplier has advised TUGCo of an iteni inIofving the possibility for cartain failures in engineered safety features actuation systems to remain undetected.

As a result of earlier concerns, specifically 10CFR50.55(e) report, SDAR CP-82-09, the supplier reconrnended corrective actions which were adopted by TUGCo. These measures consisted of special plant specific tests to.make previously undetectable failures of the P-4 Permissive detectable.

The P-4 Permissive is provided by electrical contacts in the reactor trip breakers. When the breaker is open (reactor tripped), P-4 permits the operator to block actuation of the Safety Injection System and to enter the recirculation mode. The design did not provide for on-line testing of the P-4 contacts and failure of those contacts to perform properly was undetectable.

The recommended tests resolved this by entering the switchgear cabinets and using a meter to measure the condition of the P-4 contacts.

The NRC and several utilities requested Westinghouse to give consideration to a hardware change to permit verification of P-4 without the need to enter the switchgear cabinets with portable test equipment. A change was developed and l

offered by Westinghouse as an option. This change consisted of mounting a meter and a multiposition switch on each cabinet door. It was accepted for use in fifteen plants-(including CPSES) then in various stages of construction.

It has been recently determined, after evaluating an earlier identified concern as to the overall effectiveness of the change, that the possibility of undetect-I able failures remained under certain circumstances.

Safety Implications In the event the deficiency had remained undetected, the potential exists for an undetectable loss of safeguards actuation devices.

Corrective Action In accordance with Westinghouse reconinendations, CPSES test procedures will be j

revised to incorporate the previously furnished CPSES specific testing requirements.

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TXX-4465 l

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,e Mr. D.R. Hunter, Chief Reactor Project Branch 2 U.S. Nuclear Regulatory Connission Office of Inspection & Enforcement 611 Ryan Plaza Drive, Suite 1000 Docket flo. : 50-446 Arlington, TX 76012 COMAflCHE PEAK STEAM ELECTRIC STATION UNAUTHORIZED SUPPORT REPAIRS QA FILE:

CP-85-14, SDAR-174 F1LE 110.: 10110

Dear Mr. Hunter:

In accordance with 10CFR50.55(e), we are submitting the enclosed written report of, actions taken to correct a deficiency regarding an uninvolved employee that identified some unauthorized repairs made to a hanger. This deficiency was originally reported to your Mr. D. Hunnicutt on Marcn 22, 1985.-

Supporting documentation is available at the CPSES site for your inspector's review.

Very truly yours, J.R. Wells Director, Quality Assurance JRW:tig Attachment cc: NRC Region IV - (0 + 1 copy)

Director, inspection & [nforcement (15 copies) g N N

U.S. Nuclear Regulatory Commission Washington, DC 20555 l

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ATTACHMENT UNAUTHORIZED SUPPORT REPAIRS Description As a result of'an' unsolicited observation made'by an. employee to TUGCo Management, unauthorized and undocumented welding. repairs have been identified.

Evaluation by site personnel has concluded the. concern involves.a pipe -

support ~and two (2) cable tray supports. Analysis by site Engineering has concluded the structural-integrity of the pipe support and one (1) cable tray support is not impaired as a result of the repair, The remaining cable tray support, however, will. require disposition.

The unauthorized repair involves base metal defects of indeterminate quality.

The cable tray. support provides assurance that Class IE cable installations remain. intact.during a seismic event. Failure of this support could result in overstressed'or. failure conditions for the associated electrical systems.

The affected supports have been documented by site nonconformance reports.

A recent survey by. site construction management has revealed the concern.is isolated..

Safety Implications.

In the event the defects had; remained undetected,-operator actions could be adversely affected during accident conditions.

Corrective Action in accordance with the nonconformance report, the support has been dispositioned for rework. The affected support member will be replaced. These activities should be completed no later than May 1, 1985.

Please note, the process documents involved in this deficiency are as follows:

A) Pipe Support FW-2-096-432-C62R, NCR-N-15,965, and B) Cable Tray. Supports CTH-11549 & 9760, NCR-N-85-200233 6

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UNITED STATES i ' ' ' ( 'Y E NUCLEAR REGULATORY COMMISSION i

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Washington, D.C. 20666 5,,,

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INSPECTION AND ENFORCEMENT MANUAL DEPER INSPECTION PROCEDURE 92700 ONSITEFOLLOWUPOFWRITTENREPORTSOFkNROUTINE EVENTS AT POWER REACTOR, FACILITIES PROGRAM APPLICABILITY: 2512, 2513, 2514, 252.5, 2525 92700-01 INSPECTION OBJECTIVES For nonroutine events selected for onsite followup, to determine whether NRC licensees have taken corrective action (s) as stated in written reports of the events and whether responses to the events were adequate and met regulatory requirements, license conditions, and commitments.

927.00-02 INSPECTION REQUIREMENTS 02.01 Nonroutine Event Review.

Verify the nature, impact, and cause of the nonroutine event; actions taken or planned by the licensee; and.other information of significance.

'02,02 Safety of Operation.*

Based upon the review conducted in Inspection Requirement 02.01, determine:

a.

Whether the event involved activities or facility operation in violation of the Technical Specifications, license conditions, or other regulatory requirements, b.

For facilities licensed under 10 CFR 50, whether the event involved operation of the facility in a manner which constituted an unreviewed safety question, as defined in 10 CFR 50.59(a)(2).

02.03 Reactor Trips. For reactor trip events, verify that:

a.

The licensee's program was followed for insuring that unscheduled reactor shutdowns are analyzed and that a

• This section includes written event reports that affect operating facilities and written event reports of conditions or equipment defects at plants under construction that may affect safety of operating plants.

Issue Date: OR/l'4 N

i ONSITE FOLLOWUP OF WR3TTEN REPORTS OF NONROUTINE EVENTS 92700-02.03a AT POWER REACTOR FACILITIES 1'

determination was made that the reactor could be restarted safely.

b.

_The-licensee's system for recording, recalling, and l

displaying data and information to permit diagnosing the causes of unscheduled reactor shutdowns prior to restart.

was operational.

02.04 Corrective Action.

Determine whether:

i a.

Corrective actions stated in the report 'are appropriate to correct the cause of the event,' defect, or failure to comply.

I b.

Corrective actions have been completed, are in progress, or are planned as stated in the report.

.c.

For actions not yet completed, responsibility for assuring their completion has been established (where

possible, determine the estimated completion date),

d Corrective actions include consideration of generic implications to other

systems, units, or facilities for which the licensee is responsible.

02.05-Reportina Requirements.*

Ascertain that reporting requirements have been met by determining whether:

v a.

The report is timely,

accurate, and adequately describes the event, defect, or failure to comply, b.

The safety implications and significance stated in the report are consistent with details of the event, as deter-mined in Inspection Requirement 02.01.

c.

The cause,is accurately described.

d.

The licensee plans to submit a

corrected report if information included in the report is found to be significantly in error.

02.06 Licensee Review and Evaluation. Determine whether:

a.

The event was reviewed and/or evaluated as required by the condition of license, specifications, QA program, adminis-trative controls, or other requirements.

• These inspection requirements need not be completed for every event selected for followup, but ' should be inspected periodically to verify proper functioning of the licensee's administrative controls..

l Issue Date: 08/13/84 j 1

4 l

ONSITE FOLLOWUP OF WRITTEN I

REPORTS OF NONROUTINE EVENTS l

AT POWER REACTOR FACILITIES 92700-02.06b l

b.

Review and evaluation of the event:

1.

Were based on complete and factual information.

2.

Included assessment of repeated-

events, generic implications, personnel e rro'r,

and procedural adequacy.

3.

Resulted in logical con"c$usions with regard to cause, i

significance, an.d associated corrective actions.

c.

Personnel within the licensee's organization were notified of the event as required by the technical specifications, regulations, QA program, administrative controls, or other requirements.

02.07 Documentation a.

Document the onsite followup and closecut of all written reports (including those closed during in-office review per Inspection Procedure No.

90712) in an inspection-report.

If a generic concern or a need for other followup action. is identified, inform IE,

NRR, other regional

offices, or j

other licensees as appropriate.*

l b.

Document infomation about poor performance by a-licensed Operator or Senior Operator and send a separate memorandum identifying the Operator (s) to the Regional Operator Licensing Branch or.to the Operator Licensing Branch (O LB ),

NRR, with a copy to IE.

A copy of the inspection report should also be sent to OLB.

92700-03 INSPECTION GUIDANCE' 03.01 General Guidance a.

This procedure (IP 92700) applies to those nonroutine events selected during in-office review per IP 90712 for onsite followup.

The in-office review may have been done by either a region-based or resident inspector and the onsite followup may be conducted by either a region-based or resident inspector.

b.

Pursuant to IE Manual Chapter (MC) 1105, nonroutine event reports judged by the Regional Offices to require onsite followup are to be inspected in accotdance with this inspection procedure prior to closure.

The inspection j

requirements are general in nature; requiremerits that do j

"A separate manual chapter, 0970, " Generic Issues," is being prepared.

J Issue Date: 08/13/84

ONSITE FOLLOWUP OF WRITTEN REPORTS OF NONROUTINE EVENTS-h 92700-03.01b AT POWER REACTOR FACILITIES not-apply' to the particular report being reviewed should-be omitted.

c.

Reports falling' under this inspection. procedure will normally be required because of conditions of the license

{

or permit, or by ' the rules, regulations, and orders of the

. l Commission, d.

IE may request that a specific nonroutine event written report or a category of. such reports be inspected; however, it is expected that for the majoritg of nonroutine events the cognizant first line supervisor and inspector will l

determine whether or not a site inspection will 'be pe r-I f o rmed.' Onsite inspection may be performed for:

l l'.

Written followup reports of nonroutine events I

(Licensee Event Reports) required by 10 CFR 50.73.

1 2.

Written reports (Construction Deficiency Reports) l required by 10 CFR Part 50.55(e).

3.

Written reports required by 10 CFR Part 21 (Defects or Noncompliance and submitted by the facility or determined by others to be applicable' to the j

facility.

03.02-Specific Guidance a.

Inspection Requirement 02.01.

The extent to which the inspector should independently verify circumstances of the event will be a matter of judgment based on:

1.

The complexity of the event; 2.

The licensee's' past record for reliability and completeness of event reports; 3.

The nature of questions, if any, identified during in-office review of the written report and; 4.

Radiological and/or public health and Safety concerns.

b.

Inspection Requirement 02.03a.

Review the licensee's program for post-trip reviews, which was approved by NRR, as part of the post-Sales Anticipated Transient Without Scram (ATWS) generic requirements (Section 1.1 of Generic Letter 83-28); ' based upon this review, interview licensee I

personnel and review documentation and records to ascer-I tain that the program was followed.

l Issue Date: 08/13/84 l r..

l j

ONSITE FOLLOWUP OF WRITTEN REPORTS OF NONROUTINE EVENTS AT POWER REACTOR FACILITIES 92700-03.02c l

c.

Inspection Requirement 02.03b.

Review the licenste's program for diagnostic data on unscheduled reactor shut-downs, which was approved by NRR, as part of the post-Salem l

ATWS generic requirements (Section 1.2 of Generic Letter 83-28); based upon this review, examine records and inter-l view personnel to ascertain that the data were recorded

~

and used.

7 d.

Inspection Requirement 02.04 1.

Corrective actions should generally include action to eliminate the cause or to mitigate consequences, action taken to correct the specific f ault or failure (maintenance,

repair, replacement, procedure

change, special administrative control, etc.): and action taken to reduce the probability of or to prevent recurrence I

(design: change, personnel

training, maintenance prac-tice, work controls, etc.).

These aspects of correc-tive action apply in varying' degrees to a specific event; inspector judgment is needed for this determination.

2.

The inspector should request that significant changes I

in the licensee's corrective action from that stated in the report be documented in an updated. report to the Commission.

e.

Inspection Requirement 02.04c.

If the licensee has an approved procedure for maintaining a tracking system that includes a field for noting " responsibility for completing

]

the corrective action," then the inspector should determine by sampling whether such responsibility has been documented in this tracking system.

f.

Inspection Requirement 02.05.

Regulatory Guide 10.1 gives a compilation of reporting requirements for persons subject to NRC regulations.

In

addition, required content and timing of certain types of written event reports related to this inspection procedure are specified in the following references:

Operating Reactors Technical Specifications 10 CFR 50.73 10 CFR 73.71 Defects and Noncompliance 10 CFR 21.21(b)(3)

Construction Deficiencies 10 CFR 50.55(e)(3)

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ONSITE FOLLOWUP OF WRITTEN I

REPORTS OF NONROUTINE EVENTS 92700-03.02f AT POWER REACTOR FACILITIES Theft or Loss of Licensed 10 CFR 20.402 Material Radiological incidents 10 CFR 20.403 i

I Overexposure, Radiation /

10 CFR 20.405 Contamination Levels 10 CFR 71.61 j

Packaging g.

Inspection Requirement 02.06a.

Insped,tjon Procedure No.

90714 and Change Notice 83-11, dated September 13,

1983, require review of the licensee's nonroutine event reporting program when required to determine that licensee's management, administrative controls

system, and organization exist and are functioning for investigating, reporting, reviewing, and assuring corrective actions and followup of events and incidents, h,

inspection Requirement 02.06b.

If inspection activity ioentifies significant incorrect information in the report, the licensee should submit a corrected report to the NRC.

The revised report will replace the previous. report; there-fore, the revised report should be a complete entity and not contain only supplemental or revised information.

If it is determined that incorrect information was reported or if the licensee frequently reports incorrect information, enforcement action may be appropriate.

The threshold of significance of errors, including omission, above which a corrected report is

required, involves inspector judgment and should agree with considerations for citin see for failure to report.

i, inspection Requirement 02.07a.

Documentation of event review in an inspection report serves to close out the particular event report in a traceable manner.

If more than one inspection is necessary to complete event follow-up, final closecut should be reflected in the last such inspection.

Significant corrective action items of a long-term

nature, such as certain design changes, should be tracked to completion.

NOTE:

The minimum acceptable statement for documenting the closeout of a written event report should either:

(a) state that onsite inspection was performed and summarize the findings, or, (b) state that the event was closed out based on in-office review.

~

Issue Date: 08/13/84 _ _ - _ _ _ _ _ _ _ _ _.

i ONSITE FOLLOWUP OF WRITTEN REPORTS OF NONROUTINE EVENTS AT POWER REACTOR FACILITIES 97700-03.02j j.

Inspection Requirement 02.076 1.

Significant information concerning the performance of an. individual licensed Operator or Senior Operator may have a decisive role. in licensing actions of the Regional Operator Licensing Branch or the Operator Licensing Branch (OLB), NRR.

Therefore, each inspec-tor should ensure that, $he ' appropriate group is aware of those events where the. performance of a licensed Operator may affect the safety of the facility.

2.

In general, poor performance by a licensed individual is evidenced when:

(a) The individual's action clearly demonstrates inattention to duties or disregard for require-

ments, including technical specifications and operating procedures; I

(b) The individual takes (or fails to take)

.a required action that results in significant actual or potential safety consequences; (c) There is repetitive noncompliance with l

regulatory requirements; or (d) The facility

. licensee deems that the individual's removal from licensed duties is-required.

These situations parallel those in which enforcement action against the individual may be invoked (IE Information Notice No. 79-20).

I END

(

i Issue Date: '08/13/84 j

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)

1 j

TEXAS UTILITIES GENERATING COSIPANY Log # TXX-3597 I

m nam rown. o u.w.rcw "*'"

Fi1e # 10115

/

l 1

i Decenber 3,1982 aeom "l:;'a'..i.*..',' :'

Mr. John T. Collins j

Regional Administrator U. S. Nuclear Regulatory Commission Region IV RIV

/

611 Ryan Plaza Drive, Suite 1000

. A6cket Mos. 50-445/IE Bulletin 79-14 Arlington, Texas 76012 50-446/IE Bulletin 79-14

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION 1981-B3 2300 MW INSTALLATION IE BULLETIN 79-14

Dear Mr. Collins:

In compliance with EC IE B'ulletin 79-14, CPSES has initiated a program to insure that the as-built piping and support field conditions have been verified to be consistent with the latest seismic stress analysis. This letter will serve to fonnally respond to the Bulletin.

i The scope of the program has been established based upon a detailed review of Bulletin requirements and studies initiated to identify applicable piping systems and related components along with the stress analysis problems associated with the subject piping. All related documents pertaining to each stress problem have been identified and the basis for the As-Built Verification Program is firmly established.

The following defines the scope of the piping being as-built verified to satisfy requirements of the Bulletin:

MFETYCLASS SIZE f

1 All sizes 2, 3 Large bore (2-1/2" and larger).

l 2,3,5 High energy lines over one inch that were computer analyzed.

2,3,5 Designated piping, regardless of size, (up to and including the first anchor or terminal connection)thatinteractswith safety-related large bore pipe.

Note: Class 5 piping has been defined as non-nuclear safety-related lines contained in Seismic Category I structures.

1

/

RIY official File Copr l

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j I

Safety Class 2 and 3 small bore (2" and smaller) non-high energy lines -

(regardless of the analysis method used) have been excluded from the scope of the formal 79-14 verification program for the following rea sons:

-1.

These lines are analyzed after the' piping has been installed and the as-built configuration is,known.,-

2.

Support locations used in th[Nulysis are field verified prior to completion of the analysis.

3.

Deviations from the as-designed support locations are design reviewed and reconciled with the analysis.

The verification process for a typical stress problem begins by assenbling into a " package" all related documentation in the fonn of piping and support construction drawings, and support location isometrics. The documents are then field verified by site QA

. personnel. Items verified are piping configuration, pipe support

. location and function, clearances between pipe and support, valve operator orientation, and any other information necessary to perform the stress analysis. Valve weights were previously verified in the CPSES Valve Weighting Program and this information is included in the document package. The as-built verified information is forwarded to the analysis organizations for final code analysis and piping certification.

NOTE:

. A stress problem is defined as a stress analysis of a piping system or subsystem whose boundaries are defined by.the exsistence of a fabricated anchor, simulated anchor (equipment nozzle, containment penetration), or a system of supports.

In conclusion, we a're confident that the ongoing As-Built Verification Program at CPSES fully satisfies all Bulletin requirements.

Please contact this office if additional information or clarification can be provided.

Sincerely, f.Q" 'ary 7 s

R. K. G J

RJG:grr i

cc:

U. S. Nuclear Regulatory Comnission j

Office of Inspection and Enforcement

{

Division of Reactor Operations Inspection Washington, D.C.

20555

t TEXAS UTILITIES GENERATING COMPANY

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October 25, 1979 TXX-3062 Mr. Karl V. Seyfrit, Director U. S. Nuclear Regulatory Commission

/

Region IV RIV 611 Ryan Plaza Drive, Suite 1000 Docket Nos. 50-445/IE Bulletin 79-14 Arlington, Texas 76011 50-446/IE Bulletin 79-14 COMANCHE PEAK STEAM ELECTRIC STATION 1981-83 2300 MW INSTALLATION RESPONSE TO IE BULLETIN 79-14 FILE N0: 10115 l

Dear Mr. Seyfrit:

l We are currently finalizing a procedure to address the requirements of IE Bulletin 79-14. The schedule for finalizing the procedure is consistent with the construction schedule and will be ready for implementation by December 15, 1979.

Because construction is not yet sufficiently complete to support a system inspection program, we request a waiver on the 120 day reporting require-ment.

We propose to conduct our inspections and submit our reports consistent with construction progress.

Should you have any questions, please advise.

Very truly yours, R.

. Gary RJG:df l

cc:

U. S. Nuclear Regulatory Comission l

Office of Inspection and Enforcement Division of Reactor Operations Inspection Washington, D. C.

20555

l s

.i Septem6er 7,1979 -

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UNITED STATES NUCLEAR REGULATORY COMMISSION I 0FFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 IE Bulletin No. 79-14 Supplement 2 Date: September 7, 1979 hge 1 of 2 i y SEISM!:X;A.!SIST3 RAS-5EILTSAFETY-RELATEDPIIINGSYSTEMS Desc=p icn of Circumstances: IE B G e:in No. 79-14'was issued on July 2, revised on July 18, and first supplemented on August 15, 1979. The bulletin requested licensees to take certain actions to verify that seismic analyses are applicable to as-built plants. Supplement 2 provides the following additional guidance with regu d to implementation of the bulletin requirements:- Nonconfo:sances One way of satisfying the requirements of the bulletin is to inspect safety-l related piping systems against the. specific revisions of drawings which were used as input to the seismic analysis. ' Gore architect-engineers (A4) however o are recommending that their' customers inspect these systems against the latest revisions of the drawings and mark them as necessary to define the as-built configuration of the systems. There drawings are then returned to the A-E's offices for comparison by the andyst to the seismic analysis in ut. g "'geggiar$his, gyrop, ppa. li t ~ T a A ctmit%eyg.*4 e Thefirstscpplementtothebullet[o?providedguidancewithregardtoevaluation of nonconformances. That guidance is appropriate for licensees inspecting against later drawings. The licensee should assure that he is promptly notified when the A-E identifies a nonconformance, that the initial engineering judgment is completed :in two de ad that the analyticrl engineering evaluation is c e d in 30 days. 169pmeeg i N* l 1 Visual Approximations

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l Some licensees are visually estimating pipe lengths and other inspection - s elements, and have not documented which data have been obtained in that tmy. Visual estimation of dimensions is not encouraged for most measurement; bowever, where visual estimates are used, the accuracy of estination must be within toler-ance requirements. Further, in documenting the data, the licensee must specif-ically identify those data that were visually estimated. ^; 1

1 1E Bulletin No. 79-14 l Supplement 2 Date: September 7, 1979 Page 2 of 2 I ) The d Insulatio-i In many 2: as, thermal insulation interferes with inspection of pipe support j 4 details. i.e., attach =r welds, saddles, support configua;ation, etc. In some areas,.1.e presence of thermal insulation may. gesult in unacceptably large unce_m e mrA n for determination of the location M pipe supports. Where *,nal insulation obstructs inspection of support details, the insulat:.cn sheul6 be removed for inspection of a minimum of 10% of the obstructed pipe supports in both items 2 and 3 inspections. In the Item 3 response, the licensee should include a schedule for inspecting the remaining - rupports. Where necessary to determine the location of pipe supports to an accuracy within design tolerances, thermal insulation must be removed. Clearances l For exposed attachments and penetrations, licensees are expected to me_asure or estimate clearances between piping and supports, integral piping attachments (e.g., lugs and gussets) and supports, and piping and penetrations. Licensees are not expected to do any disassembly to measure clearances. S Loose Bolti Loose anchor bolts are not covered by this bulletin, but are covered by IE Bulletin No. 79-02. 1.ny loose anchor bolts identified during actions taken for t.his bulletin should be dispositioned under the requirements of IE Bulletin No. 79-02. Other loose bolts are to be treated as nonconformances if they invalidate the seismic analysis; however, torquing of bolts is not required. Difficult Access 4 s. Areas where inspections are required by the IE Bulletin, but are considered impractical even with the reactor shutdown, shonid be addressed on a case-by-case basis. Information concerning the burden of performing the inspection and .s the safety consequence of not performing the inspection should be documented by the licensee and forwarded for staff review. Schedule The schedule for the action and reporting requirements given in the IE Bulletin s as originally issued remains unchanged. Approved by GAO (R0072); clearance expires 7/31/30. Approval was given under a blanket clearance specifically for generic problems. s a = Ik

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UNITED STATES NUCLEAR REGULATORY C0tDfISSION OITICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 Supplement IE Bulletin No. 79-14 Date: August 15, 1979 Page 1 of 2 1 SEISMICANALYSISFORAS-BUILTSAFETY-RELATEDPiikMGSYSTEMS Descrip-ic: of Circumstances: IE Bulletin No. 79-14 was issued on July 2, 1979, and revised on July 18, 1979. The Bulletin requested licensees to take certain actions to verify that seismic analyses are applicable to as-built plants. This supplement to the Bulletin provides additional guidance and definition of Action Items 2 3, and 4. 4 To comply with the requests in IE Bulletin 79-14, it will be necessary for licensees to do the following: 2. Inspect Part of the Accessible Piping For each system selected by the licensee in accordance with Itesi 2 of the Bulletin, the licensee is expected to verify by physical inspection, to the extent practicable, that the inspection elements meet the acceptance criteria. In performing these inspections, the licensee is expected tc, use measuring techniques of sufficient accuracy to demonstrate that acceptance criteria are met. Were inspection elements important to the seismic analysis cannot be viewed because of thermal insulation or location of the piping, the licensee is expected to remove thermal insulation or provide access. Were physical inspection is not practicable, e.g., for valve weights and materials of construction, the license is expected to verify conformance by inspection of quality assurance records. If a nonconformance is found, the licensee is expected in accordance with Item 4 of the Bulletin to perform an evaluation of the significance of the nonconformance as rapidly as possible to determine whether or not the operability of the systen might be jeopardized during a safe shutdown earthquake as defined in the Regulations. This evaluation is expected to be done in two phases involving an initial engineering judgement (within 2 days), followed by an analytical engineering evaluation (within 30 days). Were either phase of the evaluation shows that system operability is in jeopardy, the licensee is expected to meet the applicable technical specification action statement and couplete the inspections required by Ites 2 and 3 of the Bulletin as soon as possible. The licensee must report the results of these inspections in accordance with the require-ments for content and schedule as given in Item 2 and 3 of the Balletin. l l s I 1

s. Supplement IE Bulletin No 79-14 Date: August 15, 1979 Page 2 of 2 3. Inspect Remaining Piping i The licensee is expected to inspect, as in Ites 2 above, the remaining safety-related piping systems which were seismically analyzed and to repcrt the results in accordance with the requirements for content and senedule as given in Item 3 of the Bulletin.. < j ~ Evaluate A & r---....... ess 4A. ezana Vi in on. or the ana

ion, eMiamsd;;;.7 the licensee is exp to perform the evaluation by using the same analytical technique used in the seismic analysis or by an alternate, less complex technique provided that the licensee can show that it is conservative.

If either part of the evaluation shows that the system ot perform its intended function during a design basis earthquake, g req ^ 4B. Submit Nonconformance Evaluations .r s The licenace is expected to s t'. eval ~~ .z_ and, where,,thef1f.censee <on MJ T .'~ be csna,ervatlve, submit schedules for reanalysis in accordance with Item 4B of the Bulletin or correct the noncomformances. 4C. Correct Nonconformances If the licensee elects to correct nonconformances, the licensee is expected to submit schedules and work descriptions in accordance with Item 4C of the Bulletin. 4D. Improve Qualtiy Assurance If noncomformances are identified, the licensee is expected to evaluate and improve quality assurance procedures to assure that future modifica-tions are handled efficiently. In accordance with Item 4D of the Bulletin, the licensee is expected to revise design documents and seismic analyses in a timely manner. The schedule for the action and reporting requirements given in the Bn11etin as originally issued remains unchanged. Approved by GAO (R0072); clearance expires 7/31/80. Approval was given under i ) a blanket ciesrance specifically for generic problems.}}