IR 05000302/1975019
| ML19308D589 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 01/11/1976 |
| From: | Robert Lewis, Whitt K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19308D579 | List: |
| References | |
| 50-302-75-19, NUDOCS 8003040991 | |
| Download: ML19308D589 (19) | |
Text
{{#Wiki_filter:_ _ _ _ _ _. _ _ __ _ _ _ - _ _, _ -.... -.. <
UNITED STATES , . , NUCLEAR REGULATORY COMMISSION O' REGION ll 230 PEACHTREE STR EET. N. W. SulTE SIS ATLANTA. GEO RCI A 30303 . . . IE Inspection Report No. 50-302/75-19 , Licensee: Florida Power Corporation 3201 34th Street, South Post Office Box 14042 . St. Petersburg, Florida 33733 . Facility Name:. Crystal River 3 .- Docket No.: 50-302 J License No.: CPPR-51 Category: B1 . l Location: Crystal River, Florida Type of License: B&W, PWR, 2452 Mwt Type of Inspection: Routine, Unannounced s \\s_s Dates of Inspection: December 9-12, and 16-18, 1975 ' Dates of Previous Inspection: November 18-22, 1975 Principal Inspector: K. W. Whitt, Reactor Inspector Reactor Projects Section No. 2 Reactor Operations and Nuclear Support Branch - l Accompanying Inspector: R. F. Rogers, Reactor Inspector ! Nuclear Support Section Reactor Operations and Nuclear Support Branch G. L. Troup, Radiation Specialist
Radiation Support Section , i Fuel Facility and Materials Safety Branch S. D. Ebneter, Reactor Inspector Engineering Support Section Reactor Construction and Engineering Support Branch j ' Other Accompanying Personnel: None OVJUQ d &*ts f e I
$ \\ d M %w sl . " ' '# ' 8003040 [ j . . _
_.. -.__ _ l .. -
. l' . - -3 -) , , , . . ! i . j IE Rpt. No. 50-302/75-19-2- ' > j !I/70 ' Principal Inspector: O / . K. W. Whitt, Reactor Inspector Date i ' Reactor Projects Section No. 2 i,
Reactor Operations and Nuclear Support Branch Reviewed By: [. c' <- c / '/Z4-
R. C. Lewis, Section Leader Date I , ! , Reactor Projects Section No. 2 - ! Reactor Operations and Nuclear Support Branch ! d j I
! ,J , . , ' . !- i .
l I I i
< $ i i ! , i !
! I
i i i i i
I ! ! ' , i f l . = 1prwa-ggy vm - -yw-3m-pwpgy,q,p.
,,%.v-g,,yy ,- .meep.e,.. %q-w Mm e, et-y3-e y ymt m ogy vc ,,vggye--w-g g-1myww m wp--ep.w e gwwp&wemy7NFM w---yPN-*R&8F WNv" W W"
l . . . ___.__..J - . , ', . . . . , m[b T IE Rpt. No. 50-302/75-19-3- . SUMMARY OF FINDINGS , I.
Enforcement Matters A.
Infraction . Criterion IX of Appendix B to 10 CFR 50 as implemented by e commitments set fourth in Section 1.7 of the FSAR, Quality Program, subsections 1.7.6.7.1.1 and 1.7.6.7.1.k, states - that measures shall be established to assure that non-destructive testing is controlled and accomplished using . qualified procedures in accordance with applicable codes, standards, specifications, criteria and other special - requirements.
, Contrary to this requirement, nondestructive examinations were conducted of the steam generator tubes during the first and second weeks of November, 1975, with a procedure that did not include acceptance criteria, test frequencies, or probe scan speeds and contained other inadequacies. (Details IV, paragraph 3) ( B.
Deficiencies \\ 1.
Contrary to Criterion V of Appendix B to 10 CFR 50, licensee administrative instructions concerning approval of test procedure changes and deviations were not followed as specified by Section 3.1.1 of the FSAR, during the performance of two preoperational tests in December, 1974, and January, 1975.
(Details I, paragraph 2) 2.
Contrary to the requirements of 10 CFR 50.55(e), the failure of the auxiliary building ventilation duct during preoperational testing on November 5, 1975, was not reported to the NRC.
(Details I, paragraph 3) II.
Licensee Action on Previously Identified Enforcement Matters - , There were no previously identified enforcement matters requiring resolution.
III. New Unresolved Items 75-19/1.. Temporary Procedures ' Licensee administrative instructions provide for the use of temporary procedures. Use of temporary procedures is s ( not permitted by the proposed Technical Specifications.
\\s.
(Details I, paragraph 4) .- y p- - - , .- y- ., -.y--- ,.,---r--,- y ,-, -m
. ...- - =- . _ - - a . .. - ").
/ . . , , , - , , f \\ IE Rpt. No. 50-302/75-19-4-- . l 75-19/2 Control of Temporary Modifications and Bypasses , Licensee administrativo controls do not provide instructions for controlling certain identified aspects of temporary , modifications and bypasses.
(Details II, paragraph 3) 75-19/3 Ccre Flood Tanks Functional Test e A review of the core flood tanks functional test procedure . indicates that the licensee does not plan to perform a flow rate verification test of the core flood tanks as . recommended by Regulatory Guide 1.79.
(Details II, ! parag6aph 4). --
75-19/4 Testing of Radioactive Waste Sample Lines ' l The radioactive warte system test procedures do not ' ' include verification of representative sampling or evalua-tion of the amount of plateout in sample lines.
(Details III, paragraph 2) \\ 75-19/5 Determination of Discharge Tank Volumes s_s/ ' The actual volumes of installed tanks which can be dis-charged to the environment have not been determined to provide accurate records of discharges.
(Details III, paragraph 3) 75-19/6 Evaluation of Sampling Media Collection Efficiencies The licensee has not evaluated or documented the sampling . media collector efficiencies for filters and adsorbers used in air monitors.
(Details III, paragraph 4) , 75-19/7 Management Approval of Radiation Protection' Training Program . ' The training program for radiation protection training , has not been approved by management, although personnel are receiving training and are being qualified as radiation workers under the program.
(Details III, paragraph 5)
' 75-19/8 Indications in Class 2 components Baseline inspection of main steam and feedwater piping welds revealed indications that appear to exceed those permitted by code.
(Details IV, paragraph 4) .-
.- -- __ -- --.--_ , -, _ _
____ __._ ___._ _ _. _ _ _ .~. . . l .. .. , (m IE Rpt. No. SC ' 02/75-19-5- . IV.
Status of Previously Identified Unresolved Items . ~75-2/3, Periodic Calibration of Meteorology Facility The plant procedure for the periodic calibration of the . meteorology facility has been approved and issued.
This item is closed.
(Details III, paragraph 6) ' . 75-8/8 Boundary For Radiation Control Area - Drawings for the installation of a physical barrier at_the .
boundary of the radiation control area have been prepared and approved by plant management.
This item is closedT (Details III, paragraph 7) 75-16/1 Nondestructive Examination Procedure Qualification l , Adequate documentation has been made available to verify the qualification of the liquid penetrant procedure.
This item is closed.
(Details IV, paragraph 5) V.
Nntoual Occurrences None VI.
Other Significant Findings . Project Status-Preoperational testing is approximately 33% complete as indicated by the completion of 106 of 326 tests. Approximately 94% of the systems have been turned over from construction to generation testing.
(Details I, paragraph 8) VII.
Management Interview A management interview was held on December 12, 1975, with J. Alberdi and members of his staff. The findings of the inspection . relating to preoperational testing, plant operations, and radiation , ' protection were discussed.
(Details I, II & III) A second management interview was held on December 18, 1975, with J. Alberdi and members of his staff. The findings of the inspection relating to preservice inspection were discussed.
(Details IV) a . .,.. - -- -- .- - _., . -,,,
e,-
a-- e, ,=--g
-. _ .. .. - .. . - . - - _ _ _ _. . . _.. - . . i
% -
- . . , IE Rpt. No.'50-302/75-19 I-l DETAILS.I Prepared by:. .L [/ [[ / ' . K. W. Whitt, Reactor Inspector 'Date Reactor Projects Section.2 Reactor Operations and Nuclear , Support Branch l Dates of Inspection: December 9-12, 1975 b-
- .M
/ M-Reviewed by: . . ! R. C. Lewis, Section Leader Date ' Reactor Projects Section 2 . j Reactor Operations and Nuclear i Support Branch <- ,
' l.
Individuals Contacted
Florida Power Corporation (FPC)
J. Alberdi - Project Manager
G. P. Beatty - Plant Superintendent . ) H. E. Dumas - Test Superintendent, Electrical ! E. E. Froats - Manager, Site Surveillance
J. C. Hobbs, Jr. - Manager, Generation Testing j C. E. Jackson - Construction Superintendent P. King - QA Engineer
D. W. Pedrick, IV - Compliance Engineer C. R. Pope -. Administrative Supervisor . j C. Ritter - Test Supervisor ] G. J. Walker - Manager, Field Testing j- . i 2.
Failure to Follow Administrative Instructions ] Criterion V of Appendix B to 10 CFR 50 states, in part, that activities affecting quality shall be prescribed by documented instructions, ' ' procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with thesu instructions,
procedures, or drawings.
Section 3.1.1 of the FSAR states that all activities established in the test program are structured in accordance
with Regulatory Guide 1.68 and to comply with the provisions of 10 CFR 50, Appendix B.-. The Test Program Guide (TPG) defines the FPC administrative program for carrying out the FSAR commitments in the area of preoperational testing.
The following are examples of failure to implement the established administrative instructions as required by Appendix B to 10 CFR 50.
. i
- . _ , . . -... .. .... .. . . _. ~., _
. ,m m ' ' - . . IE Rpt. No. 50-302/75-19 I-2 s I d a.
Section 2.7.1.2 of Generation Test Procedure (GTP) 1.1 of t.ie TPC states that minor changes to class I and II test procedures shall be submitted to the Test Working Group (TWG) for revie s , and approval the next working day when possible, but in no case, later than fourteen (14) days.
Contrary to the above, . (1) Field change No. 1 dated September 29, 1975, rc TP 7 1 451 8, , Electrical and Functional Test of 120V AC and '*egulated . Instrument 120V AC Systems", was not reviewed by the TWG . within 14 days.
. (2) An unnumbered field change dated November 27, 1974, which changed step 9.2.1.74 of TP 7 2 451 10," DC Power System.- Electrical Test," had not been reviewed by the TWG at the time of this inspection. Further, the manager, Generation - Testing had not approved the change.
The test results review cycle had been completed.
b.
Section 4.1.5 of GTP 1.1 of the TPG states that test procedures must be performed in the approved sequence unless the test sequence deviation report is used in accordance with test ['"'} sequence deviation procedure, GTP-4.4 to authorize the sequence (s_,/ deviation er the test procedure authorizes specific sequence options.
Contrary to the above, (1) The test' procedure log for TP 7 1 451 8 indicates that a deviation report was filed on November 15, 1974, to permit the performance of procedure step 9.2.1.2.10 prior to step 9.2.1.2.9.
This deviation report was not contained in the test data package.
(2) One addendum to TP 7 1 451 8 contained twenty field changes.
The dates contained in the test data package indicate that these changes were not performed in sequence ' and deviation reports to justify out of sequence changes were not contained in the procedure package.
Changes 1 through 7 were performed on Jar: cey ),1975; changes 8 and 9 were performed on DeceM.cr 13, 1974; changes 10, 11, 12, 19 and 20 were p C w e.' on recember 12, 1975; and changes 13 through .'A. 231 , foraed on December 11,
1975.
Failure to follow.tdsin tstdatis e instructions is considered to be a deficiency.
,7 wT
i .
. __ ,,. , ,3 m . . . IE Rpt. No. 50-30?/7"-19 I-3 V 3~. Failure to Report the Failure of the Auxiliary Building, Ventilation Duct 10 CFR 50.55(e) requires the holder of a construction permit for a
nuclear power plant to promptly notify the Commission of significant deficiencies found in the design, construction or deviation from performance * specifications.
. Contrary to the above, the failure of the auxiliary building duct an November 5,1975, was not reported.
The ducts between the , ventilation filters and the fans commenced to distort (partial collapse) at a negative pressure between 1.5 and 3.5 inches of water during a preoperational test that was being conducted in , accordance with TP 7 2 1712, " Auxiliary Building Vent Supply and ~ Exhaust Functional Test."
The procedure specified testing with 85% of the filter area blocked to simulate a calculated pressure drop ._ across the filters of six inches of water.
Failure to report the failure of the safety related ducts is considered to be a deficiency.
4.
Temporary Procedures
Section 4.2 of administrative instruction (AI) 400 describes how temporary procedures are to be approved and used. According to this instruction, temporary procedures may be issued to direct operations during testing, refueling, maintenance, and modifications; / to provide guidance in unusual situations not within the scope of the normal procedures; and to insure orderly and uniform operations for short periods when the plant, a system, or a component of a system is performing in a manner not covered by existing detailed procedures or has been modified or extended in such a manner that portions of existing procedures do not apply.
Temporary procedures are permitted only on documented approval of the responsible section engineer and the nuclear plant superintendent, followed by subcequent review by the Plant Review Committee (PRC). The proposed Technical Specifications are silent on temporary procedures. According to the proposed Technical Specifications, operational procedures must be reviewed by the PRC prior to implementation.
The inspector stated that the proposed Technical Specification, as written, do not provide for the use of temporary procedures.
A licensee representa-tive stated that temporary procedures are needed, that ANSI N 18.7 provides for their use, and that FPC will attempt to have a section added to the Technical Specifications to permit their use.
This item is designated unresolved item 75-19/1.
5.
Verification of Review and Approval of Category II Test Procedures This item was identified in IE Report No. 50-302/75-11, Details I, paragraph 7 and subsequently discussed in IE Report No. 50-302/75-13, Details I, paragraph 4.
A few of the Category II test procedures selected on a sample basis still have not been approved.
This item V will remain open until the selected procedures have been approved by the licensee and reviewed by IE, Region II.
.
___ _ _ _ _ _ _ _ - _ _. .-. - -- . - m S , . .- - IE Rpt. No. 50-302/75-19 I-4 s s.
- . 6.
Test Results Review . The results of two preoperational tests were reviewed.
The tests were " Electrical and Functional Test of 120V AC and Regulated Instrument 120V AC System" (TP 7 1 451 8) and "DC Power System - Electrical Test" (TP 7 1 451 10).
The inspector's comments are listed as examples of an item of noncompliance in Details I, paragraph 2 ' of this report.
7.
System Walkdown ' A walkdown of the decay heat removal and low pressure injection
systems to verify that the systems had been constructed according ,_ to the FSAR was started. All the systems main lines and valves were inspected and no discrepancies were identified.
The branch lines and instrumentation taps and components will be inspected.
during a future inspection.
. 8.
Proiect Status a.
System Turnover All but three of 57 systems have been turned over to generation '- testing.
All systems turned over have exceptions to be resolved.
No systems have been turned over to and accepted by plant operations.
b.
Test Status Three hundred and fif ty-two (352) test procedures have been identified. Of these, 204 have been approved.
One hundred and six (106) tests have been completed.
There are still outstanding test deficiencies to be resolved for 39 of the completed tests.
Fifty-eight test results data packages have been approved by the licensee.
. l ( h x_ /
i . }' , k -- - -. - - -
-
3 . _' - . _ /'"'N IE Rpt. No. 50-302/75-19-II-l \\v/ . DETAILS II Prepared by: d,,4 /r.[2-y[7)' R. F. Rogers, R6 actor Inspector Date Nuclear Support Section Reactor Operations and Nuclear Support-Branch - Dates of Inspection: December 9-12, 1975 Reviewedby:/-(fd.,.u A f [ 8. C. Dance, Section Leader bate - Nuclear Support Section Reactor Operations and - a Nuclear Support Branch ' .- 1.
Personnel Contacted J. C. Hobbs, Jr. - Manager, Generation Testing D. W. Pedrick,. IV - Compliance Engineer G. P. Beatty, Jr. - Nuclear Plant Superintendent P. E. Griffith - Training Coordinator D. A. Breedlove - Administrative Supervisor \\O E. E. Froats - Manager, Site Surveillance A. L. Gomez - Director, Generation Engineering . 2.
Design Change Review and Approval The licensee's administrative controls concerning the design change review and approval process were examined.
The following procedures were reviewed for conformance with ANSI N18.7-1972 " Administrative Controls For Nuclear Power Plants" and , ANSI N45.2-1971 " Quality Assurance Program Requirements For Nuclear Power Plants."
i ' CP-ll4 - Procedure for Control of Permanent Modifications, Temporary Modifications, and Deviations Generation Engineering Control Procedure Manual , Chapters 1-7 " Design Control" . Chapters 31-38 " Safety Identification and Design Input Requirements" The inspector had no further questions.
A followup inspection is planned at FPC Generation Engineering in St. Petersburg at a later date.
5V .
_.._,- -_-
a . - - - . j
m - . , ,. fs IE Rpt. No. 50-302/75-19 II-2 k ) v .
. 3.
Control of Temporary Modifications and Bypasses The inspector verified that administrative controls had been established for controlling temporary modifications and bypasses.
Compliance Procedure-ll4 was compared to the requirements of ANSI N18.7-1972 and ' Plant Technical Specifications.
Licensee procedures require that a formal log be maintained of the status of jumpers and lif ted leads and the responsibility for control of this log is defined. Within the , areas inspected, the following problem areas were identified: , . The jumper log procedure did not require the use of approved a.
procedures when performing temporary modifications.
b.
The jumper log procedure does not require a determination by.- the responsible operator that independent verification was required following placenent and/or removal of a jumper.
c.
The jumper log procedure did not assign responsibility for determining and documenting when functional testing of equip-ment is required following installation and/or removal of jumpers.
' [ Items a through c above concerning jumper log procedures are \\s_,z} unresolved and collectively constitute unresolved item 75-19/2.
The licensee stated he would review his administrative procedures on temporary modifications.
4.
Review of Core Flooding System Functional Test , The inspector reviewed the.draf t core flooding system functional test TP 71 201 3 for consistency with Regulatory Guide 1.68 - "Preoperational and Initial Startup Test Programs for Water Cooled Power Reactors," ANSI N18.7 - 1972, and Regulatory Guide 1.79 - "Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors." The following problems were noted: a) The procedure did not provide for the flow tests under cold conditions required by RG 1.79 ART C.l.c.(1).
This step
requires that the system be tested to verify that the actual flow rate is as expected for the test conditions.
b) The procedure did not provide for the flow test under hot operating conditions required by RG 1.79 ART C.l.c.(3).
This step requires testing of each accumulator injection train by decreasing plant pressure and temperature until flow is achieved.
p.
Items a and b reflect a lack of conformance to RG 1.79 in the area of j ) core flood accumulator testing and collectively constitute N- / unresolved item 75-19/3.
The licensee stated that he would adhere to RG 1.79.
- .
_ _ _ __ - , - - _ ._ , _ _ ~\\ . / . . . , h IE Rpt. No. 50-302/75-19 III-1 % . /Z!7J!7f DETAILS III Prepared by: G. L. Troup,/ Radiation Specialist Date Radiation Support Section Fuel Facility and Materials ' Safety Branch F Dates of Inspection:, embe,r 9-12, 1975 , ,, ."r c td-
d I ' Reviewed, byrA ' ate g A. F. Dibyfon, S'ection Leader D Radiatidd Support Section , - Fuel Facility and Materials .. _ Safety Branch 1.
Individuals Contacted J. Alberdi - Project Manager G. P. Beatty, Jr. - Nuclear Plant Superintendent D. W. Pedrick, IV - Compliance Engineer O J. R. Wright - Chemical and Radiation Protection Engineer
J. L. Harrison - Assistant Chemical and Radiation Protection Engineer G. D. Perkins - Health Physics Supervisor G. R. Westafer - Technical Support Engineer E. D. Yochheim - Radiochemist (B&W) L. A. Smith - Nuclear Engineer (B&W) E. E. Froats - Manager, Site Surveillance P. R. King - Auditor Dr. P. Y. Daniel - Nuclear Support Specialist J. C. Hobbs, Jr. - Manager, Generation Testing A. P. Vogt - Testing Superintendent .
G. H. Ruszala - Test Engineer 2.
Testing of Radioactive Waste Sample Lines a.
Regulatory Guide 1.68, Appendix A, paragraph 13 requires, in part, that preoperational tests of radioactive waste systems include " tests to demonstrate that samples of liquids and gases are representative of releases" and " tests to determine the amount of plateout in sample system piping." During a l discussion concerning radioactive waste system testing a licensee representative informed the inspector that functional l-tests would be performed of the chemical addition system, which includes the liquid sampling, but that verification of repre-sentative sampling or platcout evaluation are not included as n(d \\ part of the tests, . _ .._ ,.-.m,.- ,., - , r
._ - <' s ' -. , . /'~'% IE Rpt. No. 50-302/75-19 III-2
I V . b.
Licensee management acknowledged that these t2sts should be performed and stated that procedures for these tests will be . identified and issued at a later date.
3.
Determination of Discharge Tank Volumes . Discharges to the environment from the liquid and gaseous a.
' radioactive disposal systems normally will be made from the evaporator condensate storage tanks and the laundry and shower - sump for liquids and the waste gas decay tanks for gases.
In preparing the radioactive waste release permit in accordance with plant procedure RP-104, Radioactive Liquid Waste Release - Permit Procedure, the volume of waste to be discharged is " determined by multiplying the tank level by a conversion factor while gas volumes required for plant procedure RP-105, Radioactive Airborne Release Permit Procedure, are determined by multiplying the tank volume by a pressure factor.
In both. cases the true volume in a tank requires accurate information on the physical tank volume, b.
Resins in the spent resin storage tank are pumped to a resin cask provided by a contractor for off-site shipment and dis- ' posal.
The volume of resins which are pumped to the cask are determined from changes in the storage tank resin level, which ' requires information on the physical tank volume.
c.
Licensee representatives informed the inspector that the subject tanks had not been measured to establish the as-built conditions or volumes.
The inspector emphasized the need to determine the volume of the tanks in the as-installed condition as system piping capacity up to the isolation valves affects the tank volume, especially in the waste gas decay tanks, so that a review of the as-built drawings is not necessarily indicative of the true volume.
Licensee management stated-that action would be taken to verify the subject tank volumes by dimensional check.or other appropriate methods.
. 4.
Evaluation of Sampling Media Collection Efficiencies a.
Samplers in the atmospheric monitoring system, such as, RM-A1, reactor building purge duct monitor, utilize filters and charcoal cartridges to collect particulates and halogens, respectively, present in the air stream.
In reviewing proce-dures in the Counting Room Manual, such as CH-255, Gross Gamma Spectrum of Radioactive Solids (filters), the inspector noted that equations for determining the concentration of radio- ['~'} activity collected do not account for the media collection .(,, efficiency but rather appear to assume 100% efficiency.
' . -. -. - - - -- ., . - -
. _ _ _. __ - _.
.. , -- . - - . f IE Rpt. No. 50-302/75-19 III-3 \\.N~ / - ' . b.
The inspector pointed out to licensee representatives that the collection efficiencies of the sampling media generally are - not 100%, especially for the charcoal cartridges, and may vary depending on factors such as the chemical form of the material sampled and relative humidity of the air stream.
- , These factors also make the use of manufacturer's stated efficiencies for the media unreliable unless the conditions '~ of use and of test are comparable.
The inspector asked licensee representative what evaluations had been done or would be done to substantiate the collection efficiencies " used. Licensee representatives acknowledged the variation in collection efficiencies depending on several variables and stated that no evaluations had yet been done but the efficiencies would be evaluated by tests or based on infor-mation obtained from similiar facilities.
Licensee manage-ment acknowledged this item and concurred in the course of action.
5.
Management Approval of Radiation Protection Training Program FSAR Section 12.2.2.4 states, in part, "All personnel assigned to the nuclear plant will receive specialized training in radiation safety (y s\\ " Training is presently being conducted by the CL 1;try and .... \\s ! Radiation Protection Section for plant personnel to qualify them for unescorted access to the radiation control area.
However, the training program has not been approved by licensee management.
The inspector stated that, as this training is part o; an individual's qualification as a radiation worker, the program should be approved ) by management; otherwise the training is unofficial and its use for J qualification purposes is questionable.
Licensee management acknow-ledged these comaents and stated that the program would be approved as part of the over-all training procedure, AI-1200.
' 6.
Periodic Calibration of Meteorology Facility (75-2/3) a.
This item was originally discussed in IE Report No. 50-302/75-2, Details I, paragraph 4, and dealt with the lack of procedures for . the quarterly calibration of the meteorology facility by licensee personnel. Although the meteorology facility has been periodically calibrated by contractors, FSAR Section 2.3.3, onsite Meteorological Measurements Program, requires that licensee personnel perform routine maintenance and calibration on a quarterly basis.
b.
Surveillance Procedure SP-153,. Meteorological Monitoring Instru- ' mentation Calibration, was issued on December 3, 1975, to provide instructions for the quarterly calibration of the meteorology . } facility.
In reviewing the procedure the inspector noted that the J procedure is r":ked " temporary".
Licensee management advised . .- - - - -
.- - - - e
3 , - . /N IE Rpt. No. 50-302/75-19 III-4 . the inspector that the procedure was approved for use and that after an initial use, the procedure would be revised as required-and then would be submitted for Plant Review Committee review, at which time the temporary approval would be removed.
The inspector informed licensee management that he had no further . questions on the procedure or its status of approval and that this item was considered closed.
F 7.
-Boundary For The Radiation Control Area (75-8/8) . This item was originally discussed in IE Report No. 50-302/75-8, Details III, paragraph 7, and dealt with lack of a physical barrier - at the boundary of the radiation control area to delineate the " controlled and uncontrolled areas.
Licensee representatives informed the inspector that drawings had be.n prepared for the installation of a physical barrier and that the drawings had been approved by licensee management.
The inspector reviewed the drawings of the barrier and discussed the proposed access control - and personnel flow with licensee representatives.
The inspector informed licensee management that, based on the approval of the drawings, this item was closed and that the actual installation of the barrier would be inspected at a later date.
s i . S l . Ov . -p . - - - - -. -. - - ,._ e-,, c.-, rc-,, ,-. - , -. - -+a- +.
__ e ,yF . . s
~ - . 't V} [ IE Rpc. No. 50-302/75-19 IV-1 _ /n / ls
- -/ S//75 DETAILS IV Prepared by:
/. -/Wzo //// S. D. Ebneter, seactor Inspector ' Date ~ ~ Engineering Support Section Reactor Construction and Engineering ' Support Branch Dates of Inspection * December 6-18, 1975 , Reviewed by: fN > A // / [. fyr e L.' C.' Beratyd7'Section Leader 4' jDate - Engineering Support Section , Reactor Construction and Engineering Support Branch - 1.
Persons Contacted a.
Florida Power Corporation (FPC) ~ J. Alberdi - Proj ect Manager E. E. Froats - Manager, Site Quality Assurance i J. C. Clapp - Manager, Site Quality Surveillance ) D. W. Bienkowski - Mechanical Engineer
S. Johnson - Plant Engineer b.
Contractor Organizations Babcock & Wilcox Construction Company (B&W) G. Terning - Site Manager, Preservice Inspection Team 2.
Scope of Inspection ' This inspection was devoted to an audit of the preservice inspection activities and included review of records of NDE performed on Class 2 ! systems and a review of the eddy current test program applicable t'o the steam generator tubes.
The inspector also reviewed action taken on a previously identified unresolved item.
l 3.
Eddy Current Testing FPC has completed eddy current testing of over 30 percent of the steam-generator tubes as specified in Regulatory Guide 1.83.
Some of this testing was performed in late November and early December , of.1975, as part of the total preservice inspection effort and was , conducted in accordance with Surveillance Procedure SP-305, OTSG l f-~s ( Tubes Eddy Current Baseline Inspection.
SP-305 incorporates Zetec , , . \\s_ l - l l I --. _ _ - .-
__ - . -
m - . im ( 't IE Rpt. No. 50-302/75-19 IV-2 \\s_ / ' . Procedure Z-QA-301 and the entire procedure had been reviewed and
approved by the Plant Review Committee in Meeting Number 75-17, July 24, 1975.
The inspector reviewed the procedure and found it to be unacceptable ' ' as follcus: Acceptance Criteria - No quantitative acceptance criteria is a.
contained in the procedure with regard to tube wall thinning , or through wall defects, b.
Test parameter - Significant test variables such as tape ~ recorder speeds, test frequencies and probe speeds are not_ specified in the procedure.
' Applicability - Appendix B of Z-QA-301 applies to Westinghouse c.
steam generators.
Crystal River 3 is a B&W system.
. d.
Calibration Drift - Procedure does not require re-test of tubes examined during an interval wherein calibration has drifted.
O(,) An FPC site QA review of this procedure identified and documented unspecified tape recorder speed and the inapplicability of Appendix B.
, However, no measures had been taken to correct the procedure.
The deficient procedure was used to conduct the tube examinations in early December 1975, and it therefore appears that the licensee is in noncompliance with Criterion EK of Appendix B to 10 CFR 50
and FPC implementing de:uments.
Crystal River 3 FSAR, Sections 1.7.6.7.11 and 1.7.6.7.
'.c state that measures shall be established to assure thac nondestructive testing is controlled and accomplished using qualified procedures in accordance with applicable codes, standards, specifications, criteria and other special requirements.
Contrary to these commitments, the eddy current examinations were conducted in accordance with Procedure SP-305 which is inadequate
- as detailed in the above paragraphs.
This infraction is identified as noncompliance 75-19-A1(II).
4.
Class 2 Systems The licensee is committed to perform preservice inspections on Class 2 systems per FSAR Section 15.4.4 and Technical Specification Section 4.4.10.1 both of which reference ASME Section XI, Winter 1972 addenda as the applicable code.
Section XI, Winter 1972 v _ _. -__ -_ .._ __ _ ____ _ - _ ._ ..- , A ] . l ( } IE Rpt. No.' 50-302/75-19 IV-3 v . addenda, Articles ISC, titled Rules for Inservice Inspection of , Nuclear Power Plant Components defines the inspection requirements and states-that it is applicable to systems constructed in accordance with Subsection NC of ASME Section III. Articles ISC further state ' that it is the responsibility of the owner or his agent to determine the appropriate code class subject to the review of regulatory authorities.
r FPC and its contractor, Gilbert Associates h.tvo. classified the , systems in accordance with Regulatory Guide 1. 26 and FPC in con-junction with the baseline contractor developed the preservice inspection plan around these classifications.
The main steam lines ano feed water lines from the steam generator to the outer most.- isolation valves are considered as quality group B (equivalent to i
ASME Code Class NC).
RG 1.26 and the FPC preservice inspection has classified both of these systems correctly as Class 2.
However, - these systems were fabricated and erected to B31.1, 1968, Power: Piping Code because the system design and fabrication preceded the issuance of ASME Section XI and RG 1.26.
. During the preservice inspection numerous indications were found by ['~') ultrasonic examinations in most of the weld joints in these systems.
g / As an example the data recorded on data sheet C2.1.14 for an elbcw , to pipe weld on MSSA shows 16 indications which exceed ASME Section ' XI, Winter 1972 addenda requirements.
Review of radiographic film shows correlation of some of these indications.
This situation is typical of most of the weld joints in feedwater and main steam lines.
It has not been determined that the indications are actually defects and the possibility exists that they may be due to the internal geometry of the welds.
All of the indications were detected by 45 shear wave examination and most of the reflectors are at or i-near the root of the weld.
' Florida Power Corporation is performing an engineering evaluation , of the indications but will not co'mplete it until after hydro-statically testing the system. A problem exists with regard to ' acceptance criteria.
The systems were constructed to the B31.1, 1968 code but the preservice inspection is conducted to Section XI, 1972 Winter Addenda; this latter code is more stringent in defining code rejectable defects.
This will only be pertinent if the evalu-ation determines that the indications are from rejectable conditions.
This situation'is identified as unresolved item 75-19/8 until the
engineering evaluation is complete and reviewed by NRC.
5.
Penetrant Inspection Procedure f\\ -IE Report Number 75-16 identified unresolved item 75-16/1 Non-As destructive Examination Procedure Qualification as lack of adequate m , . - ._ . . , -., _. -. - - -,. - -- -
-. _, _ _ _ _ _ _ _ _ _ _ ._.
. _. _ _. . .. _ _.. _. _ ... _, _. , _ _... _. .. ._ .. .
'N - .- , ,
l IE Rpt. No. 50-302/75-19 IV-4
-
l documentation to cubstantiate the qualification of liquid penetrant, examination procedure.
B&W had obtained adequate documentation which
listed. applicable variables and was traceable to ISI-220. This item is i closed.
. i-i F .
. i i f - ) i.
i ! I . I i i
i.
l i ! i-t i e !. E -. = -.. ... . - . _ . _, _ _ _ _ _. -. -. _. _.. _. _.. _. _.. _ _. . . . . - - - -. - }}