IR 05000302/1976011
| ML19308D346 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/20/1976 |
| From: | Jape F, Robert Lewis, Rogers R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19308D335 | List: |
| References | |
| 50-302-76-11, NUDOCS 8002270803 | |
| Download: ML19308D346 (19) | |
Text
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- g NUCLEAR REGULATORY COMMISSION
j REGION ll . f 330 PEACHTR EE STREET, M. W. SUITE St 8
% ATLANTA.SEoRGIA 30303 ..... IE Inspection Report.No. 50-302/76-11 ~ Licensee: Florida Power Corpcration 320134th Street, flouth P. O. Box 14042 St. Petersburg, Florida 33733 . Facility Name: Crystal River 3 Docket No.: 50-302 License No.: CPPR-51 Category:
Incation: Crystal River, Plorida . . Type of License: B&W, PWR, 2452, Mut Type of Inspection: Routine, Unannounced Datas of In'spection: June 15-18,1976 , Dates of Previous Inspection: May 26-28, 1976 '
\\ . . Principal Inspector: F. Jape, Reactor Inspector Reactor Projects Section No. 2 , ' Reactor Operations and Nuclear Support Branch Accompanying Inspectors: R. F. Rogers, Reactor Inspector Reactor Projects Section No. 2
Reactor Operations and Nuclear Support Branch (June 15-18, 1976) S. D. Ebneter, Reactor Inspector Engineering Support Section No.1 Reactor Construction and Engineering Support Branch (June 16-18, 1976) Other Accompanying Personnel: None , N b Principal Inspector: Af 6'M 4/2C F. Jape,'ReactorInspecep/ /' Date / Reactor Projects SectiorNo. 2 Reactor Operations and Nuclear Support Branch Reviewed by: [. d.
M M [- R. C. Lewis, Chief Data Reactor Projects Section No. 2 s Reactor Operations and Nuclear Support Branch i l l 8 002 270 h6 l \\ .. -. .- .... - . _.
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IE Rpt. No. 50-302/76-11-2- - - SUMMARY OF FINDINGS . I.
Enforcement Matters Deficiency Contrary to 10 CFR 50, Appendix B, Criterion V, calibration dates on labels for installed instrumentation have not been posted as described in Section 7.1.1.9 of the FSAR and QA 14.11.
Instruments RW-8-PT and RW-9-PT were found with out-of-date calibration labels.
(Details III, paragraph 4) II.
Licensee Action on Previously Identified Enforcement Matters Infractions 1.
Failure to Indoctrinate Personnel (IE Report 50-302/76-8) Corrective and preventive measures, discussed in the licensee's letter, dated June 2,1976, were verified. This matter is closed.
(Details I, paragraph 5) 2.
Failure to Follow Procedures (IE Report 50-302/76-8) Corrective and preventive measures, described in the licensee's letter, dated June 2,1976, were avamined and verified. This matter is closed.
(Details I, paragraph 5) ! III. New Unresolved Items None IV.
Status of Previously Identified Unresolved Items 75-8/2 Safeguards Systems Pump Runcut - The licensee is continuing to evaluate methods to prevent pump runout fa the high and low pressure injection systems.
This itsa remains open.
(Details II, paragraph 9) 75-15/1 Maintenance
Implementation of maintenance activities was verified.
Item is closed.
(Details I, paragraph 6) 75-15/7 Receipt, Storage and Handling of Equipment and Materials Instructions and implementation of safety-related items were verified. Item is closed.
(Details I, paragraph 7) , . ! I ! . _ . _. _ _ - -. . . . . _ _
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, IE Rpt. No. 50-302/76-11-3- - .. 75-16/2 Training completion and Documentation . Amendment 48 to the FSAR did not address this concern as anticipated. This item remains open.
(Details II, paragraph 5) 75-16/3 Containment Integrated Leak Rate Test Amendment 48 to the FSAR corrected disagreements be'.veen the FSAR and the test procedure. This item is closed.
(Details II, paragraph 7) 75-19/2 9,ntrol of Temporary Modifications and Bvpasses
Procedures for the control of jumpers have been revised.
' This iten is closed.
(Details II, paragraph 6) 75-19/8 InNications in class Y components ' ~ ~ ~ The licensee completed the metallurgical inspection of the sample weld and performed a borescope inspection of the internal surface of a weld. This item is closed.
_. _Detafis III, paragraph 3)., . ( 76-2/2 Instrument calibration Program This iten has been upgraded to a noncompliance of the deficiency category.
(Details III, paragraph 4) .76-3/2 Fire Water Diesel System Plant operating procedures have been revised to cover the __, fire water diesel system. Itan is closed.
(Details I, paragraph 2) . 76-6/1 Site Surveillance Audits Followup on responses to Site Surveillance Audits revealed responses are timely. Iten is closed.
(Details I, paragraph 4) 76-6/2 Precritical Test Program The licensee is writing an amendment to the FSAR to reflect Regulatory Guide 1.68 guidance. This item remains open. _(Details,II, paragraph 3) v . ,, ... _ _ _ . - _,.. _ - -. _. _.,,
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> %*. . J IE Ept. No. 50-302/76-n-4- _ . -- 76-6/3 Post Fuel Load CRD Trip Test This test procedure has been revised to incorporate
Regulatory Guide 1.68 guidance on rod testing. This item ' is closed.
(Details II, paragraph 4) ~~ 76-9/1 Diesel Cenerator Control Panels The licenses has initiated appropriate corrective action and work is in process. This item remains open until completion of the work and an audit of other class IE , > . electrical equipment enclosures is completed.
(Details III, paragraph 5) V.
Unusual Occurrences None ., _, _ _ _ _ ,,
,, VI.
Other Significant Findings Followup on IEB's 1.
IEB 76-02, " Relay Coil Failures - GE Type HFA, HGA, HKA and HMA Relays" Licensee's response, dated April 22, 1976, stated that five relays of the type described in the bulletin are in use at CR-3.
An engineering change notice has been prepared to replace the nylon coil spool with Lexan to correct the problem.
Replacement work has not been completed. Follovup of the re-placement work will be conducted on a subsequent inspection.
2.
IEB 76-03, "GE Type STD Relays" . ' Licensee's report, dated April 22, 1976, stated that these type relays are not in use, nor planned for use at CR-3.
No l further effort is planned for this bulletin.
3.
IEB 76-05. " Relay Failures - Westinghouse BFD Relays" Licensee's report, dated June 11, 1976, stated that this type relay is not in use, nor planned for use at CR-3.
No further effort isyplanned for this bulletin.
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' . . _ . .i i.E Rpt. No. 50-302/76-11-5- . - .i . VII. Manasement Interview A management interview was held on June 18 with G. P. Beatty, Jr., i and other FPC personnel.
The upgrading of unresolved item 76-2/2 to a deficiency was discussed. Followup on previous items of noncompliance, reported in IE report 50-302/76-8, and unresolved items were discussed. Also, updated status on several unresolved items, which remain open, was discussed.
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%- . IE Rpt. No. 50-302/76-11 'I-l . M DETAILS I Prepared by: 4M /C<--
F. Jape, Reactor I4sptc' tor Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch Dates of Inspection: June 15-18, 1976 Reviewed by: [. 6.
7!24'/7[ R. C. Lewis, Chief ' Date Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch 1.
Individuals Contacted Florida Power Corporation (FPC) G. P. Beatty, Jr. - Nuclear Plant Superintendent W. R. Nichols - Operations Supervisor T. C. Luckehaus - Maintenance Engineer
G. R. Westafer - Technical Support Engineer D. H. Ruzic - Results Engineer W. M. Matthews - Nuclear Shif t Supervisor R. E. Odom - Mechanical Engineer C. G. Goering - Compliance Auditor
D. W. Pedrick, IV - Compliance Engineer
J. F. Heilman - Nuclear Operator J. Lander - Construction Mechanical Engineer W. M. Lambert - Engineer J. Mack - Plant Engineer 1. E. Rogers - Manager, Stores Operation E. E. Proats - Manager, Site Surveillance , J. A. Jonee (JM) J. R. Amudson - Construction QC Gilbert Associates, Incorporated (CAI) S. Hunt - Engineer 2.
Review of Operating Procedures Procedures for the operation of plant structures, systems and com-ponents were reviewed using the guidance of R.G. 1.33 and ANSI t [ N18.7-1972 for format and content, and the proposed technical l . . i
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... . % % .. . l IE Rpt. No. 50-302/76-11 I-2 , specifications (TS), for limits and conditions. The inspector's
. review resulted in comments primarily related to inconsistencies with the proposed TS.
Licenses management stated that the pro- , cedures reviewed by the inspector and all other OP's issued prior to 1976 are being reviewed to ensure the technical limits and conditions are consistent with the proposed TS.
Licensee management could not estimate when this review would be completed, except that the review would be completed prior to the request for an operating license.
, The procedures reviewed are listed below along with the inspector's , cosaments: I ' a.
OP 417. " Containment Operating Procedure" ' Item 4.1 states a reactor temperature limit of 135 F which is inconsistent with the limit of 130 F in TS 3.6.1.5.
i Item 4.9 requires a weekly surveillance check of the air lock which is inconsistent with 72 hour schedule of TS 4.6.1.3.
Item 4.8's allows 12 hours to correct an overrun of the pressure limits which is inconsistent to the corrective action period in TS 3.6.1.4.
Item 7.6.2 requires only one air lock to be functioning which is not wholly consistent with TS 3.6.1.3.
b.
OP 605. "Feedwater System" The operating procedure refers to the safety-related portion of the feedwater system as the Emergency Feedwater System, which is inconsistent with the TS which refers to the Auxiliary Feedvater System.
, c.
OP 603. " Condensate System" Item 4.1 presents a mininnna amount of condensate dich is inconsistent with TS 3.7.1.3.
In IE Inspection Report 50-302/76-3, conuments on several operating procedures were presented. Followup on these procedures revealed that the comments on OP 207, " Fire Protection Systems," OP 305, " Reactivity B41ance Calculation," and OP 706, " Emergency Power - Diesel Generator," have been corrected and the inspector had no further questions on these ops.
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'IE apt. No. 50-302/76-11 I-3 - ' i Comments on OP 707, " Receiving Diesel Fuel," were identified as - Unresolved Item 76-3/2. Revision 3 of this procedure, issued May 17, 1976, was reviewed and all previous comments have been ' resolved. This unresolved item is closed.
3.
Fire Prevention and Protection A review of work control procedures for the prevention of and protection against fires was previously conducted and comments were , reported in IE Reports 50-302/76-3 and 50-302/76-6. Followup on these comments revealed that all concerns have been incorporated in Revision 1, dated April 13, 1976, of CP 118, " Fire Prevention Work Permit Procedure."
, CP 118 now prohibits issuance"o'f an"open flama permit during modes 1, 2 and 3 unless authorized by the Nuclear Plant Superintendent.
Also, the procedure now specifically states that the maintenance , , foreman shall approve each permit and the fire watch shall notify ! the nuclear control operator when the pern.it is.in.use.
) 4.
Site Surveillance , ' A review of audits conducted by the Site Surveillance organization j was reported in IE Report 50-302/76-6.
In this report unresolved , , , item 76-6/1 identified that corporate QA requirements did not require site surveillance audit reports to be responded to in a timely manner. Follovup on this unresolved item revealed that QAP 9, "Qualf.ty Program Implementation Audit," was revised on May 19, 1976, to require timely response to audits and QF 18.11. "FPC Quality Program Compliance Audits - Design and Construction," , requires the audited organization to implement any discrepancies noted during the audit.
The audits, reported in IE Report 50-302/76-6 that had not been answered were reexamined and responses have been r'aceived and where
l possible the audit findings resolved and closed.
Several recent audits were reviewed to determine if responses are
timely. Results of this review are as follows: Audit Number Audit Data Closure Date 327 4-8-76 5-19-76 j 329 4-15-76 .6-10-76 l 333 4-22-76 5-26-76
335 4-22-76 5-21-76 I 338 4-30-76 5-19-76 l 341 5-11-76 5-19-76 . l ' l , l 1-_,_.. . ._. - .. _. .,... .. . . ... ...... .. . . . . -. . ... - . _ _ _, _ _ _. _. -....... _,. _.
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. . % = . . , IE Ept. No. 50-302/76-11 I-4 , - Based on the findings during this inspection, unresolved item'76-6/1
is closed.
5.
Previously Reported Enforcement Matters . Licensee's corrective actions described in a letter dated June 2, 1976, regarding the noncompliance items, reported in IE Report .. 50-302/76-8, Details I, paragraph 2 were awa=4ned during this inspection and the inspector found full compliance with commitments as stated in the letter.
. Training programs were conducted to indoctrinate personnel with established QA practices. Attendance records were examined which revealed that maintenance personnel participated in the training on May 6 and 13, 1976. Technical Support personnel participated in training on May 25, 1976. Also, interviews with licensee personnel were conducted to verify the fulfillment of the coimaitment to indoctrinate personnel.
The outline and scope of the training classes were examined. These were found to cover all salient points of the FPC QA program.
I Repair and restoration records for sensing line 3A-PT1 were examined.
The line has been restored to the original condition. Weld repair was transferred to construction personnel through issuance of a W-97
form, Rework No.1311.
The inspector verified that the changes made to maintenance procedure MP 115 have been reviewed by the Plant Review Committee and approved by the Nuclear Plant Superintendent.
. Instructions for establishing and controlling clean rooms were discussed with licenses personnel. The commitment to label clean rooms with signs that state the condition of a clean room was verified. This subject was included in the traixiing classes held , '
with licensee personnel.
The inspector had no further questions regarding these infractions ' and tiiis matter is closed.
6.
Maintenance j An unresolved itsa, 75-15/1, was identified in IE Report l 50-302/75-15 and followup on this item was reported in IE Report ~ 50-302/76-3. Additional followup was conducted during this inspec-tion and the item is closed. Implementation of the maintenance ' . . .
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Maintenance." This unresolved item is closed.
) 7.
Receint. Storage and Handling Unresolved item 75-15/7 was identified in IE Report 50-302/75-15 and followup on this item was reported in IE Report 50-302/76-3.
Additional followup was conducted during this inspection and the , item is closed.
Discussions were held with Production Stores personnel and several safety-related items recently processed, utilizing the administra-tive instructions for receipt, storage and handling were reviewed.
, No discrepancies were identified. This unresolved item is closed.
8.
Plant Tour The inspector toured the reactor buidling, auxiliary building and the stores warehouse. No discrepancies were noted. During the tour,theplantwasinheatupppseofhotfunctionaltestingwith
the primary system at about 200 F and 400 psig. No significant fluid leak or piping vibrations were observed in the toured areas.
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. - IE Rpt. No. 50-302/76-11 II-1 ' . k (* 7//1[76
I DETAILS II Prepared by: '
R. F. Rogers, IQactor Inspector Date Reactor Projects Section No. 2 Reactor Operations and Nuclear ] Support Branch Dates of Inspection: June 15-18, 1976 & ?lZ$l$, Reviewed by: ?b- ' Data ) R. C. Lewis, Chief . Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch , 1.
Personnel Contacted J. Alberdi - Project Manager C. P. Beatty - Plant Superintendent J. C. Hobbs, Jr. - Manager Generation Testing D. W. Pedrick, IV - Compliance Engineer E. E. Froats - Manager, Site Surveillance R. S. Dorrie - Quality Engineer , D. Bienkowski - Engineer P. E. Griffith - Training Coordinator E. T. Childress - Startup Test Engineer 2.
Review of Zero Power Physics Test TP 71-710-1 A licensee representative stated that the concerns over the limiting values for measured moderator temperature coefficient and control rod withdrawal rates in Section 10.2 had been appropriately revised in a draft revision. This concern was initin11y identified in Report 76-7.
The inspector will review the approved procedure on a
subsequent inspection. This item raramine open.
3.
Precritical Test Program ' The scope of testing planned prior to initial criticality was compared with the requirements of Regulatory Guide 1.68 Appendix A, Section 3.1, " Precritical Tests - After Fuel Loading", and FSAR Table 13-2, " Post Fueling - Precriticality Test Summary." This concern was initially identified as Unresolved Item 76-6/2. The anticipated test program has been revised to agree with RG 1.68.
The licensee is amending the FSAR to reflect RG 1.68 guidance.
This item will r===in open pending NRC approval of the FSAR amendment.
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. IE Rpt. No. 50-302/76-11 II-2 _ ..
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Post Fuel Load CRD Trip Test TP 71-710-3 An approved copy of the procedure was reviewed for conformance to Regulatory Guide 1.68 Appendix C, " Preparation of Procedures."
This concern was initially identified as Unresolved. Item 76-6/3.
Steps involving rod drop measurements and test prerequisites have - -been revised. This item is closed.
5.
Training Completion and Documentation Operating staff training and documentation were reviewed to determine . if the training program described in FSAR Section 12.2 was properly documented. This concern was initially identified as Unresolved Item 75-16/2. Training files still require updating to reflect training required by the FSAR. A licensee representative stated that an amendment to the FSAR would be submitted which would delete training commitments for plant staff members. This item will remain open.
6.
Control of Temporary Modifications and Bypasses Administrative procedures were reviewed to determine if jumper log < controls had been revised to address the use of approved procedures and independent verification. These concerns were initially identified as Unresolved Item 75-19/2. Compliance Procedure 114, "?rocedure for Control of Permanent Modifications, Temporary Modifications, and Deviations," had been revised to incorporate these changes.
This item is closed.
7.
Review of Containment Integrated Leak Rate TP 71- %u-2 The inspector reviewed this test to determine if it reflected FSAR requirements on==imum test pressure Pa and test " instrumentation reliability. These concerns were initially identified as Unresolved Item 75-16/3. Amendment 48 to the FSAR resolved these concerns.
This item is closed.
8.
Review of Flooding Protection Measures The inspector reviewed the implementation status of commitments made in,the Safety Evaluation Report (SER), Sections 2.4.2 and 7.9.2, regarding plant physical protection and shutdown in event of ' hurricane conditions and prevention of safeguard panel flooding in
event of a fire main rupture. Required actions had been cotapleted i with the exception of installing a preaction valve in the fire main. The valve installation was in progress (ECM 2753). The inspector had no further questions.
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E Rpt. No. 50-302/76-11 II-3 " - }
- 9.
Safesuards Systems Ptano Runout The-status of the problem of safeguards systems pump runout was reviewed. This concern was initially identified as Unresolved Item f 75-8/2. A study by the A/E (Gilbert Associates) demonstrated that ' the installed orifices in each loop coupled with system pipe resist-ances preclude pump runout in the spray system. Tests of limiting flow by limiting the' motor operated valve travel through limit ! switch setpoints in the high pressure injection system were unsuc-cessful. The licensee is planning to limit flow in the high pressure injection system by throttling discharge stop check valves (MUV-2, 6 and 10). A design charge (ECN 3041) which installs permanent-locking devices for the three discharge stop check valves has been , performed. A licensee representative stated that it is still their intention to limit flow in the low pressure injection system by relying on motor operated valves throttled by limit switch setpcints.
i This item will remain open.
10.
Reactor Building Sump Flow Test TP 71-310-10 ) I Tha testing of the low pressure injection system was discussed with . licensee personnel. The adequacy of the reactor building sump ! flushing procedure was initially discussed in Report 75-8.
The initial flushing. operation for the suction line between the pumps and the sump had not assured that the lines were clean and Lees of obstruction. The reactor building sump flow test was performed on June 4, 1976 to satisfy this concern. This item will remain open pending inspector review of the completed test package.
. 11.
Test and Measuring Equipment l QA program documetLcs, calibration results, and miministrative instructions were reviewed to determine that a program for control of test and measuring equipment was implemented. The prog m was evaluated for conformance to Section 1.7 of the FSAR, Criterion XII of 10 CFR 50, Appendix B, and ANSI N45.2-1971. Approximately ten (10) surveillance test procedures performed in 1976 were reviewed for proper use and identification of test equipment. The inspector had no further questions.
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Preoperational Test Results Evaluation Comple. ad category I preoperational tests were reviewed to assure t that the licatisee had performed an adequate review of test results, the test results were within established acceptance criteria, a method of identifying and correcting test deficiencies was being . . e . . _- - _ . - . - -. .. --.. _ _ . -.
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. IE Ryt. No. 50-302/76-11 II-4 < .
) followed, and administrative control prse.tices were being followed
in test results documentation. The following emergency power ! system tests were reviewed.
TP 71-451-3, Station Battery Emergency Capacity Test CP 71-451-2, Diesel Generator Start-up Air TP 71-451-4, Emergency Diesel Gcaerator Electrical Test The inspector had no further questions.
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- .. . IE Rpt. No. 50-302/76-11 III-1 . . ' .-i fg/ :*.! l'.. ? s / / ' 7//'// 76 DETAILS III Propared by: 5 . , S. D. Ebneter, Reactor Inspector Data Engineering Support Section No. 2 , Reactor Construction and Engineering Support Lranch - - Dates of Inspec i a: June 16-18, 1976 79/7[ Reviewed by: A [[ A. R. Herde, G inf ' Date Engineering tumn Section No. 2 Reactor Coasts s!.on a'nd Engineering Support Branch > 1.
Persons Contacted . Florida Power Corporation (FPC) J. Alberdi - Project Manager G. P. Beatty, Jr. - Nuclear Plant Superintendent E. E. Froats - Manager, Site Surveillance R. C. Bonner - Electrical Construction Supervisor j D. Olsen - QA Engineer, Instrumentation K. O. Vogel - Computer and Controls Engineer G. R. Westafer - Technical Support Engineer L. Brosche - QA Engineer, NDE 2.
Scope of Inspection This inspection was devoted to a review of previously identified unresolved items. Specific items audited were Unresolved Items 75-19/8 concerning baseline data, 76-2/2 related to instrumentation calibration and 76-9/1 which pertains to diesel generator control
panel water tightness.
3.
Indications in Class 2 Components - Unresolved Item 75-19/8 The licensee has received the results of the metallurgical examina-tion performed on the pipe weld mockup. The examinations were performed by Florida Technological University for FPC and the results are documented in a report dated April 12, 1976. The examinations did not reveal any cracks or other defects that could . - be correlated with the ultrasonic test (UT) data. In addition to the metallurgical examination, the licensee visually examined the internal surface of weld MS-17A by use of a fiberoptic borescope.
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. _ _ _ - - - - _ _ _ _ - _ _ _ _ - . ~ - -, _ . IE Rpt. No. 50-302/76-11 III-2 . Entry to the internal surface was gained through MSV-93. There
were no visible machining marks in the root area of the weld that could have caused the UT indications observed during baseline ' ultrasonic eumminations. A summary of the licensee's evaluations of the ultrasonic indications follows: Numerous indications in Class 2 piping were detected during a.
preservice inspection.
~ The indications were detected with 45 shear wave and most appeared to be coming from reflectors in the root area. The pattern and characteristic of. e irdications were indicative i of geometric reflectors. No y d correlation of the UT and radiographic data could be made.
Additonal radiographs were made to a higher censitivity and b.. additional UT examinations were made on selected welds before and af tr.r hydrostatic test. Good correlation was obtained betwee-the successive UT examinations. Some variations of the UT data were also observed. However, no correlation between UT and RT data could be discerned.
A mockup of the main steam weld was made utilizing spool c.
pieces from the original pipe and the same weld procedure. A qualified welder who had welded up the main steam system during construction walded the mockup. RT and UT avaminations were performed on the mockup. The UT data again contained many indications exceeding code requirements. The RT film _ interpretation did not reveal any defective conditions that could be correlated with the UT indicationa. Careful = = 4na- _ tion of the weld root area revealed machine marks which could . have caused the indications. Preliminary metallurgical avamina-tion did not reveal any defects.
d.
Gilbert Associate (GAI) personnel evaluated the data and weld configuration. They concluded the welds were satisfactory from a metallurgical stand point and the UT indications were geometric. They s~cated that this particular weld joint design had been used at other sites and similar difficulties in NDE results interpretation had been encountered. This joint configuration is no longer used in newer generation plants.
Additional metallurgical examination was performed by Florida e.
Technological University. No cracks or abnormal conditions that could be correlated with the ultrasonic data were cetected.
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.. . . IE Ep t. No. 50-302/76-n III-3 Some of these velds will undergo subsequent ultrasonic avamina- - tion as part of the inservice inspection surveillance. The weld selection for ISI arm =4 nation per ASME Section II will be such as to detect weld degradation if it occurs. This item is closed.
4.
Instrumentation Calibration Unresolved Item 76-2/2 identified process instrumentation calibra- . tion as a potential problem. The inspector reviewed FPC documenta-i tion and procedures related to calibration of process instruments
i and those measuring and test instruments used for performing the es11brations. The FSAR Section 7.1.1.9 states "that work and quality control procedures are in general conformance with the major provisions aind requirements of the following sect 1ons of IEEE-STD-336." The list of requirements includes IEEE-STD-336, Section 6 titled Post Construction Verification.
In implementing *the calibration program, the licenses has prepared the fonowing procedures: Control of Calibration Functions j QP 12.10 - Calibration Control of Plant System Instrumentation QP 12.50 ' - Calibration Control of Portable and Laboratory QP 12.51 - and Test Equipment Use of Status Indicators QP 14.H - Quality Program Calibration and Control of Measuring Policy 12.1 - and Test Equipment Additional recuirements related to the calibration program are contained in Inter-office Correspondence dated December 19, 1975
and GTP-1.3.
The program appears to provida provisions for calibra-tion and control of instruments, measuring and test equipment and portable instrument through the plant phases of construction, preoperational test and operations although those for the latter phase have not yet been approved. It provides for traceability, recall interval, identification, repair, verification of calibration and recordkeeping. The inspector selected instrumentation in the DH system for audit to determine if procedural and FSAR requirments were being complied with. Instruments RW-8-PT and RW-9-PT were selected and the test and calibration records related to thee were reviewed. The TG-u calibration sheets and associated records were l
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__ --_ . _. _ _ _ _ _ _ . i ' . .. 'IE Apt. No. 50-302/76-11 III-4 . available. The calibration standard lab records correlated with ' test equipment entries on the string calibration sheets and the calibration lab records appeared to be consistent with procedural requirements.
The inspector audited the instrumentation in the plant and noted that the calibration stickers applied to the transmitters had not been revised to reflect the most recent calibration performed during preoperational testing. IEEE-STD-336, Section 6, Post Construction Verification, which the licenses has committed to in the FSAR, states that items requiring calibration shall be tagged or labeled on completion indicating date of calibration and iecutity of person that performed the calibration. QP 14.11, paragraph 4,9 specifies that " calibration records labels and logs are used to assure that measuring and test equipment (i.e. secondary standards, portable equipment and installed instrumentation and controls) are calibrated at prescribed intervals." Upon completion of calibration, , a label is attached to the equipment to identify the calibration ' j expiration date.
In addition, the Inter-Office Correspondence dated December 19, 1975, which delineated the Instrument Re-Calibration Program specifies that "when recalibrated, a dated sticker will be applied to each instrument." Contrary to all of the above, the calibration labels on E-8-PT and W -9-PT had dates earlier than the most recent calibration date of February 1975.
This is an apparent noncompliance with 10 CFR 50, Appendix B, Criterion V and is identified as a noncompliance of the deficiency category, 76-11-A(III). This apparent noncompliance is an upgrading of Unresolved Item 76-2/2.
The records substantiated that the transmitters had been calibrated and were acceptable. Calibration data, secondary standards records and supporting data all appeared to be satisfactory.
5.
Unresolved Item 76-9/1 Diesel Generator Control Panels The licensee has issued ECN 3342 to fabricate and install protective shields over the fan openings on ECCP-1A and EGCP-13. In addition, the' conduit entries on the top of the cabinets are being sealed. A field inspection and engineering evaluation of all Class IE electrical equipment enclosures is being performed by-GAI to identify any other areas where operation of Class IE equipment could be impaired as a result of fire protection system operation. This item remains open pending completion of the above items.
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__ _ _ _ _ . N* - I . . , . IE Rpt. No. 50-302/76-11 III-5 .l 6.
Fire Prevention Features - - The inspector inspected fire barriers and stops at various areas throughout the plant. It was noted that in the control room several fire stops had been breached due to modifications. The electrical contractor, however, had records of the breached stops which provided assurance that the stops would be repaired before final acceptance.
In addition, the inspector noted that in some instances, specifically in the regulated instrument bus panels VBDP2 and ES120 and vital bus panel VBDP4, the conduits entering the panel were not sealed.
The tire barrier constituted a seal between the floor and ceiling but fumes and gases could enter the control room via the conduits extending into the area below. The licensee stated that these conduits and others leading into the control room will be sealed.
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