IR 05000313/1974014
| ML19309D912 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 11/26/1974 |
| From: | Kidd M, Robert Lewis NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19309D893 | List: |
| References | |
| 50-313-74-14, NUDOCS 8004110742 | |
| Download: ML19309D912 (28) | |
Text
{{#Wiki_filter:. . y^g , v - - UNITED STATES C'N ' ' ATOMIC ENERGY COMMISSION ) ,l'1 ' - DIRECTORATE OF REGUIATORY OPERATIONS ,,, -l t* REGloM 11 - suite 818
230 PE ACHT REE ST REET, NoRT HWEST Tatgewegt: f 404) 930 4903
, , AT LANT A. GEORC A 30303 RO Inspection Report No. 50-313/74-14 Licensee: Arkansas Power and Light Company . Sixth and Pine Streets Pine Bluff, Arkansas 71601 Facility Name: Arkansas Nuclear One, Unit 1 Docket No.: 50-313 License No.: DPR-51 Category: B2 Location: Russellvillo, Arkansas Typ-of License: B&W, PWR, 2568 MRt Type of Inspection: Routine, Unannounced Dates of Inspection: November 5-8, and 14, 1974 Dates of Previous Inspection: October 1-4, 1974 ,
Prfncipal Inspector: M. S. Kidd, Reactor Inspector Facilities Section Facilities Test and Startup Branch Accompanying Inspectors: R. F. Rogers, Reactor Inspector Nuclear Engineering Section Facilitics Test and Startup Branch G. L. Troup, Radiation Specialist , Radiological and Envircomental Protection Branch L. L. Beratan, Senior Inspector Engineering Section , Facilities Construction Branch Other Accompanying Personnel: J. T. Sutherland, Chief Radiological and Environmental Protection Branch Principal Inspector: [, 6.
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/N M M-L- M. S. Kidd, Reactor Inspector Date Facilities Section Facilities Test and Startup Branch Reviewed By: . [. A.Lw-t b // [!" R. C. Lewis, Senior Reactor Inspector Date i i Facilities Section l j Facilities Test and Startup Branch
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. . . " ' ,. ( y .. . . A _/ ' s R0 Rpt. No. 50-313/74-14-2- ' . SUMMARY OF FINDINGS I.
Enforcement Action A.
Violations 1.
The following violation is considered to be of Category II severity: Criterion XII of Appendix B to 10 CFR 50 requires and FSAR Section 1.6.11.12 specifies that measures be established to assure that tools, gages, instruments, and other measuring and testing devices , used in activities affecting quality are properly controlled, calibrated and adjusted at specified periods to maintain accuracy within necessary limits.
Contrary to the above, the portable radiation survey instruments are not included in a program which specifies the calibration period for the instruments and control methods have not been established to control calibrated and uncalibrated instruments.
Additionally, the calibration of one type of instrument is 7ss ) not being accomplished in accordance with the written procedure.
d (Details III, paragraph 2) 2.
The following violation is considered to 'e2 of Category III severity: Paragraph 20.401 of 10 CFR 20 requires, in part, that records of personnel exposure shall be kept on Form AEC-5 in accordance with the instructions contained in that form.
i.
Contrary to the above, several Forms AEC-5 for personnel at the plant site were found not to contain information required.
(Details III, paragraph 3) II.
Licensee Action on Previous 1v Identified Enforcement Matters A.
Violations 1.
Conduct of Test Without An Approved Proceture (R0 Report No.
50-313/74-9, I.A) The licensee plans to retest the "Halon** fire contrel system using a procedure developed internally b January 1,1975.
This item remains open.
(Details I, paragraph 2) l l [ ! t j - \\s- ' l I l , % e v-+
. . . . . ,- ., fS . . \\s_- . R0 Rp.t. No. 50-313/74-14-3- , , -2.
Failure to Conduct Sampling and Analysis in Accordance with Technical Specification Recuirements (RO Report No. 50-313/74-11, I.A.2.a and b) A licensee response on this matter has been received aad corrective action and preventive measures taken have been verified. This item is considered closed.
(Details I, paragraph 3) 3.
Control of Jumpers and Bypasses (R0 Report No. 50-313/74-13, I.A.l.a) , Not inspected.
, 4.
Core Flood Tank Level An abnormal occurrence report on this item has been received and corrective actions to date verified. This item is con-sidered closed.
(Details I, paragraph 4) B.
Safety Items Oi i ( Not applicable.
. III. New Unresolved Items 74-14/1 Turbine / Reactor Trip Test , The documentation regarding inability to meet certain acceptance criteria in TP 800,14, " Turbine / Reactor Trip Test," at forty percent power was not conclusive.
(Details II, paragraph 2.a) , 74-14/2 Bad Reactivity k'orti'. Measurements The performance of TP 800.20, " Rod Reactivity Worth Measurements," had not been scheduled in the controlling procedure for power ascension testing as required by the FSAR.
(Details II, paragraph 2.b) -74-14/3 Radiation Levels Inside Reactor Containment The biological shield survey results at 40% power indicate that the radiation levels in some, areas inside of the reactor contain-ment exceed the design basis levels established in the FSAR for 100% power.
(Details III, paragraph 4) - . .
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. - - - _. .- - -. ._ . , e , - - .. , , . \\-l '. (G [RORpt..No. 50-313/74-14-4-74-14/4 Contamination and Airborne Radioactivity Survey Forms A review of contamination survey forms indicated deficiencies in,the recording of required data on the forms. Also, it was noted that the airborne radioactivity forms lack spaces for signoffs by the individuals performing surveys or analyzing the samples-or for ~a supervisory review.
(Details III, paragraph 5) 74-14/5 Regulatory Operations Bulletin 74-12 ~
This will be carried as an unresolved item pending receipt of information requested in the ROB.
' 74-14/6 Regulatory Operations Bulletin 74-13
i This will be carried as an unresolved item pending receipt of information requested in the ROB.
74-14/7 Reactor Building Spray Line '\\ On November 7,'1974, a leak was dis tovered in the suction line for spray pump P35B. Two additional leaks were discovered November 13, 1974. The licensee has committed to evaluating the cause of the cracks and determine if others might exist in other plant _ systems which utilize similar piping.
(Details V, paragraph 2) ,
) IV.
Status of Previously Identified Unresolved Items 74-11/1 Control Room Ventilation System > The licensee is preparing to test the control room normal and emergency ventilation systems to demonstrate that they meet
the qualifications given in the FSAR. This item remains open.
'(Details I, paragraph 5)
' V.
Unusual Occurrences i A.
Reactor Coolant Pump Suction Line Drain Line Leak On October 18, 1974, the licensee reported that a'small crack had been discovered in a coupling weld on the drain line for "B" RCP suction piping. The_ weld was ground out, renaired, and operations resumed. On October 26, 1974, it was rel ;ted that the repaired weld had cracked in a different locatioa. Corrective . actions are discussed in the second abnormal occurrence report, . dated November 4, 1974.
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Reactor Building Spray Line Leak ' , . On November 7,1974, licensco personnel reported that a small leak had been discovered in the 10-inch suction line for reactor building spray pump P35B. Two additional leaks in this system were discovered November 13, 1974. An abnormal occurrence report is to be submitted by November 17, 1974.
(Details V, paragraph 2) C.
Reacto,r Building Cooling Units Drains An unusual event report involving the inability of the drains for , the containment coolers to carry design flow was submitted per Technical Specification 6.12.3.2 October 11, 1974. Corrective actions are described in the referenced report.
(Unusual Event Report 50-313/74-2) D.
Pressurizer Relief Discharge Piping Hangers This unusual event was reported October 11, 1974. Several pipe hanger rods wa.re bent during a pipe shock test on the discharge / -~s line for the pressurizer electromatic relief val e.
Corrective actions are described in the licensee report.
(Unusual Event Report 50-313/74-3) - VI.
Other Sienificant Findings A.
Review of Power Ascension Testing A review of certain power ascension tests results was conducted.
(Details II, paragraph 2 and Details III, paragraph 6) B.
Regulatory Operations Bulletin 74-9 ' A followup report on findings of investigations requested in this R0B of GE Model 4KV "thgna-Blast" circuit breakers was submitted October 1, , 1974.
C.
Independent Measurements Program i The licensee's strontium analysis of the particulate filter test standard for the independent measurements capability test was in disagreement with the AEC standard. ' Another test standard has been sent to the licensee for analysis.
(Details IV, paragraph 2.c) O]s. t !
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. - . . D p~. - f~h . . I ) \\ RO Rpt. No. 50-313/74-14-6-s,e . VII. Management Interviews A.
A management interview was conducted November 7,1974, to discuss findings of inspection of the licensee's program for radiation protection and on November 8,1974, to discuss other findings of the inspection con-ducted to that date. The following representatives attended: Arkansas Power and Light Company (AP&L) J. W. Anderson - Plant Superintendent R. G. Carroll * - Chemical and Radiation Protection Engineer T. H. Cogburn** - Nuclear Engineer R. R. Culp** - Test Administrator C. A. Halbert* - Technical Support Engineer L. W. Humphrey** - Quality Assurance Engineer G. H. Miller - Assistant Plant Superintendent J. L. Orlicek - Quality Control Engineer B. A. Terwilliger** - Operations Supervisor
- Attended November 7, 1974, meeting only.
- Attended November 8, 1974, meeting only.
,,s The violations in Section I were discussed.
(Details III, paragraphs 2 and 3) - The status of previously identified violations in Section II were discussed.
(Details I, paragraphs 2-4) The new unresolved items in Section III were discussed.
(Details II, ' paragraph 2) ' The previously identified unresolved item in Section IV was discussed.
(Details I, paragraph 5) B.
J. W. Anderson, Plant Superintendent, was informed by Region II via telephone on November 13, 1974, of the results of the independent measurements program.
(Details IV, paragraph 2.c) C.
A management interview was conducted November 14, 1974, to discuss findings of the inspection of cracks in the reactor building spray system piping.
(Details V,' paragraph 2) . V . - . . .. . . -. ..
. -. .. - . -. . ' , . ^ O - . G RO Rpt. No. 50-313/74-14 I-l DETAILS I Prepared by: ,[.(. t< v b / ///,)6/8 M. S. Kidd, Reactor Inspector Date Facilities Section . Facilities Test and Startup Branch , Dates of Inspection: November 5-8, and 14, 1974 // dh-Reviewed by: M 8-
R. C. Lewis, Senior Reactor Inspector Date Facilities Section Facilities Test and Startup Branch 1.
Persons Contacted In addition to those listed in the Management Interview Section, the following persons were contacted during this inspection:
Arkansas Power and Light Company T. C. Baker - Chemical and Radiation Protection Engineer ~ Bechtel Power Corporation W. McMahon - Lead Startup Engineer J. Oszewski - Project Engineering 2.
Halon Fire Control System ' This matter involved testing of the control room "Halon" fire control
system with a procedure which had not been reviewed and approved by AP&L.
(R0 Report No. 50-313/74-13, Details I, paragraph 4.a) A licensee letter, dated November 1,1974, revealed that evalua-tion of-results of the original test by AP&L resulted in declaring them unacceptable. This will require retesting, which is to be - accomplished using a procedure developed by AP&L. Resolution of this item is' expected by January 1, 1975, according to the referenced letter. This item remains open.
I 3.
Failure to Conduct Sampling and Analysis in Accordance with Technical Specification Requirements j-This matter was initially discussed in R0 Report No. 50-313/74-11, Details I, paragraphs 4.a and 4.b.
A licensee response on this y enforcement matter, dated Octaber 1,1974, discusses corrective actions, t \\ %_,/ ' i , , , - - -,
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, - . A } D - . , O LR0 Rpt. No.'50-313/74-14 I-2 - ' steps taken to avoid future violations, and dates when full com-pliance was achieved for each instance of noncompliance with Technical Specification requirements. Verification of certain corrective actions , by the inspector were discussed in the referenced inspection report. During i c the current inspection, the inspector verified that the surveillance report i of Technical Specifications requirements described in the licensee letter of October 1, 1974, was being used to monitor fulfillment of
< . requirements by station management. Also, the completion of analyses on ' the dates specified in the letter _were verified as well as for certain ~ samples taken subsequently. Licensee management were informed that
this item was considered closed.
' 4.
Core Flood Tank Levels This abnormal occurrence was discussed in R0 Report No. 50-313/74-13, I.A.1.b and involved a violation of the limiting condition for operation'for core flood tank level as defined in Technical Specifi- ! cation 3.3.3(A). A licensee report was subnitted per Technical l Specifications on October 11, 1974. Corrective actions described ' in the licensee report were discussed.. Licensee personnel stated that the cause of loss of water from the reference -legs had been , determined to be leakage through packing on isolation valves for the-legs as evidenced by buildup of boron crystals. The monitoring , i program on the reference levels was continuing and no problems had becn experienced since the valve packing leake 77-corrected.
Licensee personnel indicated that they felt the problem had been
corrected and that no additiona1' instrumentation or changes in , instrumentation would be needed. This item is considered closed.
! ! 5.
Control Room Ventilation System
, This unresolved item was initially discussed in R0_ Report No. 50-313/74-11, . i ' ' Details I, paragraph 9.1. (10). (c). Results of preope ational tests on ' - the control room normal and emergency ventilation syst ems did not i demonstrate that they met the design bases and qualif' cations discussed in Section 9.7 of the FSAR.
During this inspection, results of testing on the control room emergency ventilation system and the computer room system were [ reviewed and. discussed with licensee personnel. These tests had
not been completed or reviewed, but' indicated that these systems ! probably would meet design bases. No testing had been attempted i .on the normal control room ventilation system because of leakage j-through backdraft dampers for each of the supply fans VSF-8A and VSF-8B.
(FSAR page 9-30a and Figure 9-13).
This leakage had re- , l' suited in flows approximately twenty percent less than expected.
, j Repair of the dampers was expected to take six to eight weeks,
after'which a performance test would be run. This item remains open.
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V R0 Rpt. No. 50-313/74-14 I-3 . 6.
Changes in Power Ascension Testing Program - The inspector discussed changes which the licensee had made in the power ascension testing program since the operating license was issued.
One group of changes was made through revision of the FSAR description of the program in Amendment 46-dated September 20, 1974 A review of Safety Review Committee -(SRC) minutes of a meeting on August 14, } 1974, revealed that the committee had reviewed these proposed changes to the FSAR and found that no "unreviewed safety questions", as defined by 10 CFR 50.59, were involved.
- Another change -involved the deletion of a turbine trip test at 40 percent power. The AEC. document, " Guide For The Planning of Initial Startup Programs," recommends that this test be run at 40 percent and 100 percent power. The AP&L cvaluation on this item noted that i
Regulatory Guide 1.68, "Preoperational and Initial Startup Test Porgrams For Water-Cooled Power Reactors," recommends that the test be run only at 100 percent power. A reactor trip test was performed at 40 percent. The inspector was assured that a turbine trip test will be performed at 100 percent. power. The inspector had no 's questions on this matter.
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( \\ - \\ / 1 R0 Rpt. No. 50-313/74-14 II-1 ///[M74 DETAILS II Prepared by: ve. ov - Lu--- R. F. Bogers( III, Reactor Inspector ' Ddte
Facilities Section .
Facilities Test and Startup Branch Dates of Inspection: November 5-8, 1974 Reviewed by: /) d.
eb OI~'d/'7d R. C. Lewis, Senior Reactor Inspector Date j Facilities Section Facilities Test and Startup Branch - 1.
Personnel Contacted
Arkansas Power and Light Company (AP&L) J. W. Anderson - Plant Superintendent ' R. R. Culp - Test Administrator f'~'s T. H. Cogburn - Nuclear Engineer I J. L. Orlicek - Quality Control Engineer (J . 2.
Review of' Test Data The inspector reviewed the results of four power ascension tes: procedures; TP 800.05 Reactivity Coefficients at Power, TP 800.14 Turbine / Reactor Trip, TP 800.20 Rod Reactivity Worth Measurements, TP 800.28 Psuedo Rod Ejection. During this review, the following discrepancies were detected: a.
TP 800.14 Turbine / Reactor Trip Test . , Two acceptance criteria were not satisfied in the performance of this test. Paragraph 8.1.01/002 requires that high pressure injection not be i initiated and Paragraph 8.2.01/003 requires that pressurizer level remain between 40 inches and 300 inches. High pressure injection was manually initiated due to decreasing pressurizer level while the actual level reached approximately 31 inches. The corrective j action listed in the test document for these deficiencies indicate , ' that none is possible and that these deficiencies are characteristic of the primary system.
.This test was approved for final acceptance by the station superintendent on Form A-16, Test Endorsement Record, on October 4, 1974. This form states that "All deficiencies and discrepancies have been cleared and all acceptance criteria have been net."
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A _ /. s , c In. disc'us,sions(with' licensee representatives, the inspector uns shown a letter from. Babcock and Wilcox dated October 29, 1974,~ which-
. provided technical justification for a lower pressurizer level limit l
' and recommended :that the. licensee revise its acceptance criteria as ' ' , ~ . presently stated;in'this test'. This had 'not been done. Inability ' . to meet test acceptance criteria in the power ascension test program - j must be fully documented'and evaluated prior to final acceptance . by the plant superintendent. This evaluation and acceptance must ' be completed'for.this test'and will remain an unresolved item.
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. . ' - b. 'TP 800.20' Rod Reactivity Worth' Measurements , ' This test is listed as required-in Enclosure 9.7 of TP 800.01,' the controlling procedure for power ascension testing.
It-is required to be performed at zero power and "at power" in Section 13, page 13-45, of the FSAR. The at power portion of this test is not , currently a requirement of the controlling procedure sequence. A
i licensee representative stated that'he had planned to perform this i test at the.75% plateau. This test needs to be formally incorporated - ! ' into the test sequence and performed as required. This item will j remain an. unresolved item.
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3.
Power Level Plateau Data Review %. Verification that the test results for the 0%,15%, and 40% plateaus were
. [ . evaluated by.the licensee _and that they were determined to be acceptable . prior to proceeding to the next plateau was performed. Test Working - Group Committee review of each plateau was satisfactorily documented.
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+ ! RO Rpt. No. 50-313/74-14 III-1 , i 7[ e _ b DETAILS III Prepared by:G. L.7rDup, Ragiation Specialist Date Reactor Facility Section, Radiological and Environmental Protection Branch Dates of Inspection: November 5-7, 1974 i 7/ Reviewed by: _Mp k _ Date' A. F. Gibsca,/SeniogHealth Physicist Reactor Facility Section, Radiological and Environmental Protection Branch 1.
Individuals Contacted J. W. Anderson, Jr. - Plant Superintendent G. H. Miller - Assistant Plant Superintendent C. A. Halbert - Technical Support Engineer R. G. Carroll - Chemical and Radiation Protection Engineer J. L. Orlicek - QC Engineer 2.
Control of Portable Radiation Survey Instruments The calibration of portable radiation survey instrunents is ,/ 'y a.
(C') accomplished in accordance with station test and inspection procedures written for each_model of instrument (Eberline Model PI;R-4, Eberline Model 20-1, etc.). The inspector re-viewed the calibration procedures for five different models of instruments and noted that none of the procedures specified a calibration frequency. During discussions, licensee repre-sentatives stated that the general practice was to calibrate instruments on a quarterly basis (i.e., every 3 months) and that recalibration was controlled by the " calibration due date" on the calibration sticker affixed to the instrument, a " tickler" file maintained by the health physics staff, and a weekly computer print-out showing which instruments were due for calibration during the upcoming week, but they acknowledged that there was no of ficial schedule for the calibration of these instruments. The inspector noted that each of these methods was based on the unofficial quarterly calibration period rather than a formal calibration program as required by 10 CFR 50, Appendix B, Criterion XII.
b.
Some portable radiation survey instruments are maintained at the access point to the controlled area for use by workers and health physics technicians while the remainder are stored in a cabinet in the health physics office for use by health physics technicians or as needed. The inspector selected four instruments at random from the ccbinet. Two instruments had no calibration stickers affixed, one instrument had a calibration sticker affixed but which indicated a period of four months between calibrations, and one instrument had I a calibration sticker af fixed but the recalibration due date had , passed by approximately two weeks. Licensee representatives stated "
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" /'% . I l i _,/ RO Rpt. No. 50-313/74-14 III-2 s . that personnel hnd been instructed not to use instruments unless a calibration sticker was af fixed so that the two instruments with no stickers would not be used. The instrument uhich indicated a four-month period betvcen calibrations was checked against the calibration records.
It was determined that the wrong date of calibration had been entered on the sticker; this was corrected by licensee representa-tives to reficct the proper calibration date on the sticker and with a three-month calibration period. Licensee representatives further stated that the normal tolerance on the recalibration period is + 25% so the instrument for which the recalibration date had passed was acceptable for use. The insmector then checked two additional instruments in the cabinet and found that both had calibration stickers affixed and that the instruments were within the calibration period on the sticker. The inspector pointed out to licensee representatives the potential problems associated with the nixing of calibrated and uncalibrated instruments in the same cabinet.
Criterion XII to Appendix B to 10 CFR 50 requires that measures be established to assure that tools, gages and instruments are properly controlled.
On November 7, 1974, an AEC inspector exited the controlled area at c.
the designated access control point. While frisking himself, he ['__sj noted that the calibration sticker affixed to the frisker ratemeter ( / (Eberline ibdel R11-14) had a recalibration due date of August 18, 1974 A licensee representative told,the inspector that someone had removed ,, the ratemeter which had previously been installed and installed this ratemeter in its place that day. The ratemeter was removed from the access control point and checked for being within the calibration accuracy for the instrument. A licensee representativ.e informed the inspector that the check indicated that the instrument was within the calibration accuracy prescribed for that type of instrument.
d.
The inspector reviewed the calibration procedures and records , for five different models of instruments used in the Biological Shields Survey (TP 361.01). The records indicated that four of the models of instruments had been calibrated in accordance with the procedures and were with the prescribed limits for accuracy.
Paragraphs 8.2.3-8.2.5 of Station Test and Inspection Procedure 1303.72, Teletector Calibration, required that the Eberline Model 6112 Teletector be calibrated in 0.75 mR/hr fic1d. A review of the calibration records for Teletector serial number 55391 indicated that the instrument has been calibrated in a 0.5 mR/hr field for the last two calibrations. The calibration records for the other Teletector instrument were also reviewed. Those records indicated that a 0.5 mR/hr ficid was used in the past tro calibra-tions. A licensee representative stated that the change in the field strength was made due to the difficulty in obtaining a 0.75 mR/hr field using the calibrator at the plant and acknowledged - ~: ( ) that the procedure should have been revised accordingly.
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.2 .5M f A T ' [)e s. a . (~_ j)' ' - RO Rpt. No. 50-313/74-14 - III-3 . Licensee' management acknowledged.the above examples of lack of e.- control over portable radiation survey instruments and stated that the following corrective action would be taken: , (1) A recalibration frequency would be officially establi'shed ' i for the survey instruments and would be included in the. calibration procedure or in the station procedure i for Operational Test Control (OP 1004.12)
. - (2) The separation of calibrated and uncalibrated instruments l
in' storage would be considered.
t (3)LThe calibration procedure for the Teletector instruments.
+ would be revised to reflect-the current calibration technique.
, _3.
Completion of Form AEC-5 ,a.
Personne : 2diation exposures were recorded on Form AEC-5, " Current l Occup'ational External Radiation Exposure," as required by Paragraph j 20.401 of 10 CFR 20.
Paragraph 20.401 requires that records-be j-kept on the Form AEC-5 in accordance with the instructions in that-form. The inspector reviewed the current Forms AEC-5 for personnel , _
at the plant site and noted several discrepancies between the form and the requirements contained in the form. Types of discrepancies noted were as follows (not all descrepancies noted on same form): '
(1) only individual's last name entered vice full name ,
- (2) neither the Social Security No. nor "none" not entered (3) whole body dose status not entered (4) method _ of monitoring for neutron exposures not indicated al-though neutron exposures were recorded on the form.
b.
Licensee management stated that all Forms AEC-5 would be , reviewed and all required information which was missing would ! be included on the forms. Additionally, licensee management stated that personnel would be instructed to devote more attention to assuring that these records were completely and
~ properly filled out for each person.
- ! 4.
Radiation Levels Inside Reactor Containment , a.' FSAR Figures 11-4 through ll-9 define the design basis radiation ' j zones in-the plant. Areas within the reac*or containment are defined -as = Zone 'lV (< 100 mR/hr) or Zone V (>100 mR/hr). Zone IV ' - are those regions outside of-the secondary shiald which are accessible lto personnel with the reactor at 1Gu% power.
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R0 Rpt. No. 50-313/74 1.4 III-4 \\ < , . b.
The recalts of the Biological Shields Survey (TP 361.01) at 15% and 40% power indicate that some areas at elevations 336, 357 and 401-which are classified in the FSAR as Zone IV have radiation levels in excess of 100 mR/hr, which exceeds the design basis for 100% power.
c.
Licensee management stated that this condition is' currently being reviewed by the architect-engineer (A&E) who designed the shielding and that the A&E is evaluating possible modifications of the shielding and/or installation of additional shielding. Licensee management also stated that corrective action to be initiated by AP&L is deferred pending completion of the A&E's design evaluation.
5.
Contamination and Airborne Radioactivity Survey Forms a.
The inspec : reviewed radiation and contamination survey records for the month of October and noted a general deficiency in the completion of the data required by the survey forms.
Examples of problems encountered are (although not all are found on each survey form and not all forms had errors): (a) data entered in the wrong columns (N) disintegrations was not entered on the sheet although (b) the instrument conversion factor for converting counts to g the factor was used in eq1culations (c) the instrument used in analyzing contamination survey swipes was not entered (d) counting time, and/or background count rate were not recorded although values for these factors were used in calculations.
t.
b.
In the review of the survey records the inspector noted three forms which. contained computational errors. These were identified to a licensee representative who had the records corrected immediately. The inspector had no further questions on these survey forms.
Each day the instrument conversion factor for converting counts c.
to disintegrations is determined by the health physics staff for use in the analysis of contamination surveys. Once determined, this factor is recorded on the survey form and used in calculating the amount of radioactivity present on the surfaces surveyed. At the time of the inspection this conversion factor was simply entered on a slip of paper which was then taped to the counting instrument ("d/c = 4.2") but neither the date nor the i sme of the person doing the determination was entered on the paper. A standard form was __, / available for tais determination which included all of the information (,)) relevant to the determination and allowed the determination for each day during a month to be r~ corded on one sheet and any trends identified.
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. . . < . A licensee rep: asentative stated that this form had pre.iously been ' used for recording of the data. During the inspection licensee representatives initiated the use of the form for recording the.
daily determinations. The inspector had no further questions on this , l - matter.
d.
The form for recording radiation and contamination surveys is provided with spaces far the recording of the name or initials , of the individual.who performed a survey, analyzed the swipes and reviewed' (audited) the completed form. The inspector noted that the form for recording airborne radioactivity surveys had
none of those spaces for the recording of the names or initials so.that it was impossible to identify the individuals who were involved in the survey from the records.
Licensee management stated that the following corrective action ^ e.
would be taken: (1) review existing survey procedures and evaluate the need for additional instructions or guidance; make changes as appropriate e (2) revise the airborne radioactivity survey form to include spaces for identification of individuals conducting and reviewing the survey , (3) reemphasize to all health physics personnel the need to ' properly fill out survey forms.
6.
Biological Shields Survey ' i The inspector reviewed Test Procedure 361.01, Biological Shields Survey, and the results for the background, low power and ' intermediate power surveys. Several instances were noted where - .the completed survey records were in disagreement with the require-
ments of the procedure (instrument type and/or serial number , ~ recorded, wrong instrument used for survey and improperly or not ' .the results of the reverification survey with the proper instrument not recorded, data not entered on data sheets and no notation to indicate that the particular survey had been deleted, etc.). Licensee management stated that the survey results would be reviewed again and-the missing data would be entered when possible to reflect the actual , ~ t performance of the surveys.
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, . / / A i \\- (,/ R0 Rpt. No. 50-313/74-14 IV-1 DETAILS IV' Prepared by: e b 0 / N ~ W. L. Britz, Radiat' ion Specialist Date Radiological and Environmental Protection Branch ' ' Dates of Inspection: October 9,1974 Reviewed by: 2%
- ~N~7[
A. F. "Cibson, vSenior Health Date Physicist, Reactor Facility
Section, Radiological and Environmental Protection Branch 1.
Individuals Contacted J. W. Anderson - Plant Superintendent R. Carroll - Chemical and Radiation Protection Engineer, ANO-1 J. Holman - General Of fice Chemist, Little Rock, Arkansas 2.
Indeoendent Measurements The licensee is required to measure the quantities and a.
concentrations of radioactive material released in effluents N - I as a result of operating his facility.
This requirement is to
- - !
assure and to demonstrate cqmpliance with the limits specified in AEC Regulations and the limits that are specified in his operating license.
b.
This inspection consisted of tests of the licensee's capability to measure radioactive material in test standards which approximate samples of actual radioactive effluent. The tests consisted of com-
paring the licensee's measurements with those of the AEC's reference , ' laboratory which prepared the tect standards. The measurements made by the AEC's laboratory are referenced directly to the National Bureau of Standards radioactivity measurements system by laboratory intercomparisons.
The results of the inspection are given in Table 1.
The criteria c.
used for the comparisons are given in the attachment to these details.
Licensee analyses were performed by the ANO-1 laboratory and also the AP&L environmental laboratory at Little Rock, Arkansas. The tests were in agreement except for the strontium measurements on the filter paper which were in disagreement. The strontium measurements were performed by the AP&L environmental laboratory because the ANO-1 laboratory was not ' capable of performing the measurements at that time. The ANO-1 laboratory now has the capability for strontium measurements and a test standard has been sent to the ANO-1 laboratory for their analysis. The environmental laboratory has been sent another test standard to check ['~'\\. their strontium measurements.
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.,. .- -. . . ' - g j"*) I i \\j CRITERIA FOR C0" PARING ANALYTICAL MEASURE!ENTS General The following provides criteria for comparing results of capability tests. The criteria are based on an empirical relationship which combines prior experience'and_the accuracy needs of this program.
In these criteria, the agreement limits vary in relation to the ratio of the AEC Reference Laboratory's value to its associated uncertainty. As that ratio, referred to in this program as " Resolution", increases the acceptability of a licensee's measurement should be more selective.
Conversely, poorer-agreement must be considered acceptable as the resolution decreases.
Criteria LICENSEE VALUE RATIO (-- ) AEC REFERENCE VALUE < Possible Possible Resolution Agreement Agreement A Agreement B <3 0.4 - 2.5 0.3 - 3.0 No Comparison .) s,/ 4-7 0.5 - 2.0 0.4 - 2.5 0.3 - 3.0 , 8 - 15 0.6 - 1.66 0.5 - 2.0 0.4 - 2.5 16 - 50 0.75 - 1.33 0.6 - 1.66 0.5 - 2.0
-200 0.80 - 1.25 0.75 - 1.33 0.6 - 1.66 >200 0.85 - 1.18 0.80 - 1.25 0.75 - 1.33 , "A" criteria are applied to the following analyses: Gamma Spectrometry where principal gamma energy used for
identification is greater than 250 Kev.
i Tritium analyses of liquid samples.
Iodine on adsorbers.
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V-2- . "B" criteria are applied to the following analyses: ' Gamma spectrgmetry where principal gamma energy used for- ' identification is less than'250 Kev.
, , 89 Sr and 90 Sr Determinations.
Gross Beta where samples are counted on the same date using the same reference nuclide.
Procedure a.
The AEC Reference Laboratory value should be divided by its l associated uncertainty (la) to obtain the resolution.
b.
The ratio of the two measurements to be compared should be ! determined by dividing the result to be compared by the AEC Reference Laboratory result.
! I Agreement is considered obtained if the ratio falls within c.
the range given in the "Agreenent" column for the aggociated ' resolution. For example, consider a comparison of Sr determinations. A liransee obtains a value of 1.97 +.08 x 10-5 . (y,,h/ uCi/ml and the AEC Reference Laboratory reports a result of 2.53 +.06 x 10-5 uC1/mi. The resolution would be 42, i.e. 2.53/.06, , and the ratio is 0.78, i.e. 1.97/2.53. This pair of measurements would be considered to be in " agreement" because for this > resolution, the " agreement" range is 0.75 - 1.33.
d.
If " agreement" is not achieved, the ratio should be evaluated for "possible agreement".
In this case, consideration is =ade for the type of analyses conducted by selecting a range in the appropriate column; i.e., "A" criteria or "B" criteria.
If the ratio falls outside the appropriate "possible agreement" e.
column, the two measurements will be considered to be in ' " disagreement".
f.
Licensee results are Nul ACCEPTABLE for isotopes that are not ' identified by the licensee but are identified by the AEC reference lab as being present in concentrations greater than 107. of their l respective !TC's as specified in 10 CFR 20, Table II.
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_ _ _ _ _ _ _ _ _ _ _._ _ _ - _. }* Uffluent X ' .. , Er.viror. mental , AEC-Licensec Sampic Labcratory Doing Tested u . Verification Capability X
- -
AP&L AN Analyzed HSL Results UM M Results E555 Eampic Descriptio:: For uCi/nl uCi/ml Liquid, #3 H-3 1.1 i.02 E-3 1.24 E-3 1.27 , I!n 54 2.88 i.019 E-3 2.76 E-3 2.56 Fe 65 5.1 i.14E-3 4.74 E-3 4.56 Cc-144 5.65 i.23 E-3 4.97 E-3 5.29 . Sr 89 4.8 i.05 E-3 5.30 E-3
T' Sr 90 1.31 i.028 E-3 1.07 E-3
Filter Cc-144 5.23 i.37 E-3 4.45 i E-3 4.86 Cs 137 2.40 t.037 E-3 2.12 E-3 2.07 Co 60 4.3 1 093 E-3 3.87 E-3 3.72 Sr 89 0.736 +.014 E-3 , - 0.215 E-3 e , , Sr 90 0.404 i.0037 E-3 0.135 E-3 - - - - Liquid B-1 Gross 6 2.31 1.023.E-3 2.22 E-3 2.77 ,
O O . .
. . . s D O) (Q R0 Rpt. No. 50-313/74-14 V-1 . d in k. // 2*T- ? 4 DETAILS V Prepared By:. . L. L. Beratan, Senior Inspector Date Engineering Section Facilities Construction Branch Dates of Inspection: November 14,'1974 ,; _.
',/ ' Reviewed By: - .' C. E. Murphy, Chief Date i Facilities Construction Branch 1.
Persons Contacted . Arkansas Power and Light Company (AP&L) H. Miller - Assistant Plant Superintendent J. Orlicek - Quality Assurance Engineer 2.
Scope of Insnection The purpose 'of this inspection was to examine the cracks in the heat- ) affected zone of several of the welds in the 10 inch Schedule 10 stainless ~~ / steel pipe. This pipe is located.on the suction side of the reactor building spray pump, and a cross-over line between two containment spray systems.
Also to determine what measures would be taken to examine other systems using this type of material.
3.
Observations Two of the three cracked sections were on the suction side of the reactor building spray pump. Each crack was adjacent to a weld in the heat- ' affected zone.
The cracks were intermittent and circumferential.
In each case a repair had been made to the weld during the assembly of the system. A third crack was found in a cross-over line between two valves but near a 90 elbow. The material certifications identify these pipes as P10 and P15 and that in the elbow as E20. Co ies of c the material certifications are attached.
4.
Examin: tion of Cracked Pipes Crack No. 1 is adjacent.to the weld in the heat-affected zone where the pipe passed through a concrete block, wall. A section of this cracked pipe was sent to Bechtel - San Francisco, for a metallurgical examination.
Preliminary results indicate that cracking was intergranular and . appears to have been a result of stress corrosion.
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. . - - . /, , ~~ . .( ) R0 Rpt. No. 50-313/74-14 V-2 w The pipe with crack No. 2 was still at the site and when examined by the inspector the cracks were found to be adjacent to the weld in tha 5 to 6 o' clock position. The cracking was circumferential. A specimen of - this pipe was requested so that the AEC could make an independent examination of the failure.
It was agreed by the licensee's management that a specimen approximately 1-1/2 to 2 inches wide and extending approximately 3 inches on either sides of the weld would be furnished.
It was agreed that a similar specimen would be given to Bechtel for examination.
5.
Examination of Schedule 10 Piping Systems . To determine if there are additional defects in the schedule 10 pipe a UT examination program was proposed, and it was planned to examine all the welds in the affected cystems. Because of the thinness of the pipe UT examination did not prove to be effective and this scheme was abandoned. A program of RT examination would be pursued. By November 15, 1974, five jointe had been RT'd and no defects were detected; but it was reported to the inspector on November 17, 1974, that 17 joints had been RT'd and one showed some indications on the ID of the pipe. Visual gs examination of this joint will be made on a daily basis until it can g j be repaired.
V 6.
Other Systems With Schedule 10 Pipe A program of examinction of welds and the heat-affected zone of all schedule 10 stainless steel pipe where ever used is being considered and the RO:II ' will be kept informed of what type of program will be adopted. This will remain an unresolved inspection item.
7.
Lopairs - The removal of the cracked sections 1 and 2 was completed at the time of the examination. The cross-over line had not been campletely drained so that this-portion of pipe had not been cut out.
The pipe with crack No. 2 was cut off at the face of a check yalve. The valve was dressed back to prepare it for welding in a new section of pipe. Upon examination it appeared that either all of the old weld had not been removed or a sliver of old pipe remains attached to the valve casting.
The licensee stated that they would examine the valvo preparation and determine if additional material must be removed before the pipe weld is made.
All repairs are being made in accordance with Bechtel Corporation General Welding Standard, GWS-SN, Revision 1, dated May 1, 1971, and Welding Procedure Specification P8-T-Ag, Revision 9, dated March 5, 1971.
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