ML20087F806

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Compliance Rept 50-219/69-02 on 690209-0404.Major Areas Inspected:Status of Resolution of Items Previously Listed as Outstanding in Insp Rept 50-219/69-01
ML20087F806
Person / Time
Site: Oyster Creek
Issue date: 05/28/1969
From: Robert Carlson, Gilbert R, Nolan F
US ATOMIC ENERGY COMMISSION (AEC)
To:
Shared Package
ML093631134 List: ... further results
References
50-219-69-02, 50-219-69-2, NUDOCS 9210120100
Download: ML20087F806 (43)
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{{#Wiki_filter:n-3=v _..w I, yn, g,T c a,)Seu yi c / U. S. ATOMIC ENERGY COMMISSION REGION I i DIVISION OF COMPLIANCE Report of Inspection R CO Report No. 219/69-2 y ; s? ? Licensee: JERSEY CENTRAL POWER & LIGHT COMPANY (Oyster Creek 1)

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Construction Permit No. CPPR-15 l Category B D:tes of Inspections: February 9 - 13, 16 - 20, 24 - 28, 1969 j i.] March 12 - 14, 19 - 20, 24 - 27, 1969 April 2 - 4, 1969 j. ,1 ') Dates of Previous Inspection: February 3, 4 and 6, 1969 j dAl9 h 6/n/M In:pected by: R. T. Carlson, Senior Reactor Inspector bat'e f ));! (Responsible Inspector) I l l. Oad.~ YS r/zde F. J. o an, Senior Reactor Inspection Specialist 'Date j f 69 R. G. Gilber't, Radiation Specialist Daife ' Reviewed by : h 5!At!69 .j

f. G. K'e#Ver, Senior Reactor Inspection Date Specialist l

Proprietary Information: None f ,l I ,~. y

SUMMARY

43' The Division of compliance's pre-licensing inspection program for Jersey Central Power & Light Company's Oyster Creek 1 reactor facility 1 uce completed on April 4, 1969. This report documents the status of resolution of the items previously licted as being outstanding in the summary section of CO Report No. 1 219/69-1. The report also speaks to the status of two issues which 4

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6 .~ ,y ~~ v. i t. w..) I developed subsequent to the inspections covered by the referenced f report, and which also required review before a recommendation s could be made by Compliance to the Division of Reactor Licensing [ regarding the issuance of the Provisional Operating License. One of the latter issues pertains to the Compliance investigation into allegations concerning the adequacy of certain pipes, fittings and valves used in important systems in this facility. The other item relates to a question regarding the main steam flow re-strictors. A summary report of the above information was submitted to Headquarters by TWX on April 8, 1969. Certain of the items, including the areas under investigation, were identified in the TWX as requiring additional review by Compliance. However, because they were of such a nature that they were not important to the safety of initial fuel loading and low power testing operations, it was recommended that a limited Provisional Operating License be issued. It was stated that the limiting conditions should prchibit reactor operation at significant power until the accept-ability for nuclear service of important pipes, fittings and valves have been substantiated. Provisional Operating License No. DPR-16 was issued to Jersey Central Power & Light Company on April 9, 1969. The license limits the initial operation of the facility to 5 Mwt and without the reactor head in place, to permit initial fuel loading and testing pending (1) modification of the standby gas treatment system, (2) additional review of the quality of certain piping in the facility, and (3) evaluation of preoperational testing of contain-ment isolation valves.

g o. -- __n n. .z 1, a.! w .), I. Scope of Visits The compliance pre-licensing inspection program for Jersey { Central Power & Light Company's Oyster Creek 1 reactor facility was centinued to completion. Inspection visits were performed during j the period covered by this report as follows: February 11-13, 19-20, 24-26, 1969 Carlson (section II) April 2-4, 1969 Nolan (Section III) February 9-12, 16-19, 24-28, 1969 March 12-14, 19-20, 24-27, 1969 March 12 and 13, 1969 Gilbert (Section IV) Technical assistance was provided Mr. Carlson during the February 11-13, 1969 visit by Mr. W. J. Collins, Metallurgist, j CO:HQ, Mr. R. M. Gustafson, Materials & Metallurgy Branch, DRS, and Dr. R. G. Gilliland, Parameter, Inc. (CO Consultant). l Technical assistance was provided Mr.. Gilbert during the March 12 and 13, 1969, visit by Mr. L. D. Denton, Inspection Specialist (Health Physicist), CO:HQ and Mr. O. L. Cordes, Chief Health Physicist, Phillips Petroleum Co., NRTS-Idaho (CO Consultant). The site was also visited by Mr. R. T. Dodds, Reactor Inspector, CO:V, on February 11-13, 1969, for orientation preparatory to his inspection involvement scheduled to commence with initial fuel loading. Mr. D. L. Caphton, Reactor Inspector, CO:I, accompanied Mr. Dodds during the February 13, 1969 visit. Mr. R. T. Dodds and Dr. R. G. Gilliland visited the GE-APED Laboratories, San Jose, California, on February 20, 1969, to review the results of GE's metallurgical examination of a defective valve (Paragraph II.A.l. of report). The principal persons contacted during these visits were as follows: Visits to Site Jersey Central Power & Light Co. (JC) G. H. Pitter, Vice President T. J. hcCluskey, Plant Superintendent D. E. Hetrick, Operations Supervisor

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4-G. 'l i I. R. Finfrock, ' Jr., Technical Supervisor l D. A. Ross, Technical Supervisor (In-Training) i N. M. Nelson, Maintenance Supervisor

k R. D. Doyle, Chemical Supervisor.

{ D. E. Kaulback, Radiation Protection Supervisor R. Sullivan, Assistant Technical Engineer .- t ~I V. D. Thomas, Instrumentation. Engineer (GPU Nuclear Group) 1 E.J. Riggle, Electrical and Instrument Foreman ~ ' ".. 'I 1 General Electric Company-(GE) R. L. Dickeman, Manager, Domestic Turnkey Projects (San Jose) J. Barnard, Manager, Licensing Activities - Domestic Turnkey Projects (San Jose) L. M. Loeb, Manager, Materials and Quality Services, Domestic Turnkey Projects (San Jose) D. K. Willett, Project Manager (San Jose) R. C. Christenson, Manager, Plant Test Engineering, Power Plant Projects, APED (San Jose) R. A. Edggins, Principal Project Engineer (San Jose) F. Brutschy, Manager, Field Engineering Chemistry (San Jose) R. Osborne, Chemist, Field Engineering Chemistry (San Jose) J. W. Eberle, Consulting Engineer - Balance of Plant Engineering (San Jose) K. W. Hess, Site Operations Manager W. C. Bibb, Operations Superintendent D. W. Diefenderfer, Principal Test Design & Analysis Engineer i N. M. Strand, Site Construction Manager R. M. Haynes, Lead Construction Engineer D. L. Borchers, Electrical Construction Engineer J. C. Larrew, Test Engineer J. Staley,-Test Engineer Bu rns and Roe, Inc. (B&R) G. A. Lari, Project Engineer (Oradell) i MPR Associates, Inc. (JC Consultants) I T. E. McSpadden, Metallurgical Engineer i i i 2 4 --r l . i

e ( t / , 1 II. Results_ of Visits - Carlson f A. Status of Outstanding Issues f The status, as of the last visit covered by this report, gj of those issues identified in the summary section of CO Report No. .( 219/69-1 as requiring resolution before a recommendation could be j made to DRL relating to the issuance of the Provisional Operating a License, is summarized below. For purposes of comparison, the paragraph numbers below correspond to the item numbers in the referenced report. 1. Valve Crack Problems The following additional facts regarding GE's program for investigation and correction of the subject problems were established: a. Valve V-14-34, a 10" isolation valve (outuide dry-well) in the isolation condenser system condensate return required replacement. This was accomplished using a comparable valve (Crane) originally scheduled to be installed in Dresden 2. One complication was that it required conversion from an AC valve operator to one powered by DC. This was completed and the related preoperational testing satisfactorily completed. b. Valve V-16-2, a 6" isolation valve (outside drywell) in the cleanup demineralizer system supply, was also found to be sufficiently defective to require replacements. As in the case of valve V-20-17*, the defects were principally shrink cracks detected through the review of the vendor 2adiographs. The defects were confirmed by reshooting in the field. GE proposed to complete the valve replacement subsequent to initial fuel loading but prior to power ascension. Co found this proposal to be acceptable. c. Valve V-20-17, core spray system (inside drywell), was successfully repaired in place. At one time GE thought the valve might require replacement due to the nature of the defects.**

  • CO Report No. 219/69-1, Paragraph II.A.4.d.
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e q ) 4 d. Per Messrs. Loeb and Strand, all valves that required replacement or repairs possibly affecting the j integrity of the pressure barrier were rehydro'd 4 per applicable code requirements. These included l the following: ]) (1) V-17-19 (2) V-17-54 (3) V-20-17 (4) V-20-34 (5) V-20-41 e. The procedures and techniques of weld repair of the Anchor valve casting defects, and subsequent non-destructive testing, were reviewed by the inspector with the assistance of Dr. Gilliland, Mr. Collins and Mr. Gustafson, during the February 11-13, 1969 visit, and found to be proper and correct. f. The following observations were made by Dr. Gilliland during the February 20, 1969 visit to the GE-APED Laboratories (made to review GE metallurgical examina-tion of valve V-20-41):* (1) "The metallographic examination showed the casting defects to be the result of solidifica-r tion shrinkage and hot tearing, and were manifested as linked microvoids. i (2) The examination revealed no excessive mold-metal surface reactions. (3) Pre-and post-heat treatment radiography indi-cated that the defects were truly casting imperfections, and that the heat treatment sequence did not propogate the defects. "

  • Excerpted from full report of visit.

Latter available for review in Region I office.

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GE. (Messrs; Loeb and Tackett) attributed the Anchor g. I valve problems to the lack of specificity ~in! purchase-specifications regarding the timing of radiography,- i L together with poor quality control including radio-' - !~ ' graphic interpretation at the first line vendor. level) with a lack of'second or. third level followup (B&R,. l '4 GE).' CO concurs in this evaluation' j m i 3 h. Repair-of_the defects.found in the' recirculation-system pump suction valves. (Chapman) in loops'.B and. j 4 C* was successfully completed. This'was accomplished: by first removing'the defects by grinding to a depth l of approximately 3/4" (~1/3 the wall thickness), and. j then welding back to the minimum wall; thickness nl followed by dye penetrant inspection. CO found'the above repair procedure to be acceptable.- 1. During the February 11-13, 1969 visit, Mr. Collins l reviewed the before and after radiography of a-1i previously questioned weld _ joining one set of the-special double isolation valves in the isolation i condenser system.**.The_ review showed that GE had q reshot the wrong weld. The correct weld was reshot j during the visit,and Mr. Collins confirmed that'the: i ~' indications in question were indeed film artifacts. This matter is considered : resolved. I In telecons with Messrs.'McCluskey,:Hess and Loeb d subsequent to the February 20, 1969 visit to the GE 'l laboratories in San Jose, the inspector. informed ' those l persons that CO concurred in their. evaluation as to the j cause of.the crack problems _ experienced.with the' valves supplied by Anchor, and in their investigation and s repair program as far as it went. However, since it had (f been found neceussary to make_ repairs of'some sort'in such ~ a large percentage ( ~ 25%) of-the: valves inspected, and-j K ~ d.- since the problems detected in at least' two~ of the valves (V-20-17 and V-16-2) came to light principally-as the- ]j result of GE's review of original vendor _ radiography, it was CO's position that this latter review =should be expanded to include all stainless steel Anchor valves j in systems that communicate with ' the primary system.. j q i

  • CO Report No. 219/69-1, Paragraph II.A.4.h.

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    • CO Report No. 219/68-10, Paragraph'II.A.17.

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g ) -8.- vi 's O They were informed also that.the review need not be completed before the initiation of fuel loading; however, i that it. should be completed in altimely. manner., i' i This matter was still under consideration by'JC-GE at. 4 the time of the April 2-4, 1969 visit. Issue not considered to be an item affecting issuance of the -I operating license, i .2. Cable Tray Loading GE provided CO'with a copy of a summary report

  • of their investigation in this area, including the mock-up tests conducted at Bridgeport, ' Connecticut.

Some discrepancies were noted between both the assumptions and results in the report and those presented orally by GE at the December 8, 1968, ACRS meeting, attended by the writer. Further, the test results indicated that actual temperatures likely to be experienced would be higher than originally predicted. It was noted that GE i did not propose to conduct any in-plant verification' measurements. The results of CO's review in this area, including further discussion on the ' subject in a DRL-CO meeting with JC-GE at Headquarters on February 19, 1969,** was that CO was not able to arrive at a position of reasonable assurance on this matter without some verifica-tion, by actual measurement, of the results of the GE study. GE subsequently agreed to conduct a program for monitoring cable temperatures during plant operation; however, the details of this program had not been made i available to CO at the time of the April 2-4, 1969 visit. Issue not considered to be an item affecting issuance of the operating license. 3. Calibration of Effluent Monitors .i The results of CO's continued review in this subject 1 area are discussed in Section IV of this report.

  • Letter from J. W. Eberle to R. A. Huggins, Oyster Creek Cable Tray Thermal Analysis, dated February 10, 1969.

Copy available in Region I files.

    • Minutes of meeting with JCP&L Co., from Tedesco to Boyd,' dated March 6, 1969.

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a ~w* j. ., ) r As a result of observations made, JC-GE were informed -) by this-inspector that the location of the filters in [ the stack sampling line raised serious doubts that [ these filters will see concentrations of halogens and 4 particulates that are representative of those being f released from the stack. JC-GE then agreed on a calibra-1 tion program that will include sampling of halogens and N particulates at the probe sampling point to demonstrate the adequacy of this system. JC-GE (Messrs. McCluskey and'Hess) stated that should the system' prove inadequate, 7 appropriate modifications will be made.. On this basis, this matter is c'onsidered to have been adequately resolved and not considered to be an iten affecting issuance of the operating license. 4. Containment Penetration Restraints This matter was referred ' to DRL for resolution. The i subject was included amongst those discussed at the previously referenced February 19, 1969,. meeting held at i Headquarters, and is also spoken to in Amendment Nos. 50 i and 51. DRL's eventual finding of adequacy on this matter is reflected in their Addendum to the. Safety Evaluation, dated April 9, 1969. l 5. Nondestructive Testing of Safety' Valves At the time of the last visit covered by this report, JC-GE were developing a proposal which they stated would be responsive to Regulatory concerns. Issue not considered to be an item affecting issuance of the license. i I. 6. Preoperational Testing j l The results of CO's continued review in this subject area l are discussed in Section III of this report. In summary, those tests previously agreed to by CO as requiring completion prior to initial fuel loading

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found to have been satisfactorily completed with the following exceptions:

  • CO Report No. 219/69-1, Paragraph II.H.

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._ to _ ,( [ y .c Test No. A-1,:Drywell and Absorption' System-Leak-a.- Rate Measurement. This. test.was observed to be l[7 - demonstrating the capability of the sys' tem to conformi satisfactorily completed with the. exception of. 1 s-to-a statement '. in the Application ' (Amendment L No.1 11', 4' Section III, Question 10,.page III-lO-12,. Paragraph .t .7.7) that "no valve leakage rate is to exceed 1.0% '1 M of the allowable containment leakage volume per 24-hours.-" ~ This limit was exceeded for both _ sets of rW 1 main steam isolation valves. .7 - b. Test-No..C-14, Standby Gas Treatment System and l Reactor: Building Leak' Rate' Test. This test was. not-i completed due to the inability of the system:as designed to meet the Technical Specification' require-ment of 0.25" vacuum at 1,200 cfm-air flow. l Item 6.a. was referred to DRL for resolution, as had been item 6.b. previously (see Amendment No. 52)., The disposition of these items by DRL'as it relates to the-issuance of the P rovisional Operating License is' reflected in the Notice of Issuance of-the POL, dated April.9,.:1969. j R 7. Operational Environs Monitorino Program Discussions with Messrs. McCluskey and Finfrock at the time of the April 2 - 4, 1969 visit revealed that, contrary. i to earlier indications,* JC now plans to continue'the' preoperational environs monitoring program, i,e., that- ] described in the Application,.into the operational phase of the facility. As a result, this matter is' considered-i by CO to have been adequately resolved. -The assigned T inspector will confirm the satisfactory. implementation of the subject program. o 3 -1 8. Containment Spray System Check Isolation Valves-j T.. The intent of the Application -is not clear with regard to the presence of check, valves in the containment spray system inside the drywell. Following discussions with JC-GE and with DRL**, CO considered the matter resolved.

  • CO Report No. 219/69-1, Paragraph'II.D.
    • Letter, Tedesco to Boyd, Minutes of Meeting with JCP&L Company, Docket No. 50-219, dated March 6, 1969.

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.n , W. n_s.4,_1 - ~ 'I a. 'i K r. ;' -.. '* .p n i b' [ +'q 11 - -} g. w; a l s 9. . Outstanding Construction Items 1# During the. April 2 .4,.:1969 visit, the inspector,.in thel f .t company of Messrs. Christianson, Hess and Rossi. conducted. ~ q [

a. tour lof,the. entire plant.

The status of construction l Jh.1 was observed to be in accordance with the GE proposal' s i previously. agreed to by Co.* This matter is. considered. ] ,f .to have been satisfactorily resolved.. 9 -l 10. Core Loading ' Procedure The final, JC-GE approved version of the initial core loading procedure was reviewed by the inspector.and found to have adequately resolved areas of concern - j previously identified by CO.** In addition, the' final-1 JC-GE approved revision to their Administrative' Procedure for Initial-Fuel Load and Start-up Tests was. reviewed f and ' found to have been updated to incorporate the organiza-4 tional changes' reflected in Amendment No. 52. Thisfitem- -i is considered.to have'been satisfactorily resolved.: 3 y 11. Startup Testing Program j The inspector's review showed that the resolution of: outstanding issues in this area'had been' satisfactorily 1 completed with the - exception of 'JC-GE formallyf approving - specific. test procedures relating to the post-loading -j test phase of the startup program. JC-GE (McCluskey 'I and Hess) stated that :the approvals..were scheduled to be i completed during the loading phaset Issue not considered-l to be an item-affecting issuance of ~ the -license. l ] 12. Plant Operating Procedures k The outstanding items requiring CO followup *** were f reviewed by the. inspector and found to have been resolved t in accordance with previous understandingsi reflected in the referenced report. This matter is considered to have been satisfactorily resolved.

  • CO Report No. 219/69-1, Paragraph II.H.
    • CO Report No. 219/69-1, Paragraph II.B.4.
      • CO Report No. 219/69-1, Paragraph IV.F.

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~ ..i. 13. JC-GE Operating Organizations Ni The-inspector's review of the JC-GE' organizations to. I determine conformance with the commitments in the j; Application and readiness to assume. operational 4- -responsibilities revealed.severalfsignificant-f' ' deficiencies-relating-to the numbers =ahd types of people assigned, past and pending. changes in. personnel, '(j and the qualifications-of those personnel assigned. n CO's concerns'in this area were.further. compounded by~ the failure' of certain key individuals to pass or even. 1 to take the cold operator licensing exams. j ~ These issuas were subsequently discussed with JC-GE at' a meeting with REG, held 'at Headquarters Lon February. 27, 1 1969. As a result, JC-GE submitted Amendment'52 which included additional information on the start-up organiza-- ] tions.- The referenced document outlines a proposed re-? ) . alignment of responsibilities within'the existing.organiza- ) tions together with - the addition of additional. personnelo o to augment th'e operations and-technical staffs-in certain ~ - - areas of responsibilities. Also, a second cold. operator licensing exam was conducted to satisfy the concerns of -l i REG in this area. ? DRL's eventual ~ finding of adequacy in ' this subject area l is reflected in the Notice of Issuance'of' Provisional' l operating License, dated-April 9, 1969. rl' .14. Emergency Evacuation Alarm The subject alarm was test sounded at the request of the inspector on several occasions.during the. facility tour -h discussed'in paragraph II.A.9. Areas'of remoteness-1 "l-and/or high background noise leveli were singled out. The- ' alarm,.which was. noted by the inspector to be. quite. + penetrating, was found to be clearly audible in.all instances. This item is considered to have been. satisfactorily completed. r +d 's +em w. -.. w.epe-.. e_ g,,%,_ w a vt'7 T-*--m. m.m m.

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.t 7 i ~ 15. open sleeve in-watertight Compartment Bulkhead-The inspector visually' determined'that GE'had.' subsequently l j g '"~t welded shut one end of the; subject' sleeve..This matter j i is considered to have been adequately resolved. 4 3 h 'l py d B. Status of New Items q 1 l ' The status of issues which developed subsequent to the-last' R ') . report and which also required. resolution-before a. recommendation could be made by CO' to DRL relating to the. issuance of the Provisional j operating License, are summarized below: 1. Allegations'Concerning the Adequacy of Certain Pipes, Fittings and Valves On Feburary 28, 1969, CO:I received a letter' dated February 21, 1969, from Dr. Roscoe E..Kandle,-Vice-J 4 . Chairman, New' Jersey Atomic Energy Council, reporting i that on Thursday, February 13, 1969, certain individuals-of a New Jersey piping firm had informed representatives I l of the Public Utilities Commission that-pipe had been supplied for use in OC-1 which did not meet appropriate i specifications. The results of CO's subsequent investiga-- tion, which led to additional l allegations including-some-1 pertaining to certain fittings and valves at; this facility, are documented in separate investigation reports.*L At. the time of the April 2 - 4,1969 visit, the investigation-had revealed the allegations to be. of sufficient substance-and significance to warrant the' recommendation by CO to DRL** that the Provisional Operating Licenselbe issued- ] [ with. limitations that would prohibit reactor' operation.- l at significant power and remain in effect pending i satisfactory resolution of these issues. The> facility) license, DPR-16, dated-April 9_,1969, limits-initial u I

  • CO Investigation Report, Jersey Central Power & Light Co.,-Oyster, Creek Unit No. 1, Docket No. 50-219, Interim Report Nos. 1 and 2,

-dated April 25, 1969 and May 6,-1969, respectively, and final . report (to be issuad). +

    • TWX report from Carlson to O'Reilly, d.ated April 8, 1969, and q

forwarding memorandum from Engelken to Morris, dated April 8,-1969. + ,,,w, m .,r _..,,m

p ~.... [ h ,N ^ M'" '~" m ~ ._ 14 _. [ l g-operation to 5 Mwt and.without the ' reactor head - in-place, J 1 pending resolution of' 'three issues -including that _ discussed-j ,j here.. This matter remains. outstanding attthe_ time of this report. 2. Main Steam Flow Restrictors The design of the main steam' flow.restrictors-installed-1 ] ' at this. facility was reviewed intlight of the problems ?3 L detected at Dresden.2..* The review here showed the l units to be of a different design and different manufacturer - from those at Dresden 2, - and apparently without theo problems noted. The manufacturer'of the oyster: Creek 1 -l units is BIF, Division of "the; New: York Air. Brake f Company; A copy of the related shop drawing, No. ' A-166586Jia available in the CO:I files. 1 t . III. Results of Visits - Nolan

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4 This section of the report, which is concerned with the. pre-- .i operational testing program, was prepared by Mr. F. J. Nolan.. A. Scope of Compliance Review 1 Results of the preoperational testing program were reviewed j on the dates indicated in paragraph III.C. on the.following page.. The tests that were incorporated in this review-were the specific 'l preoperational tests that were required to be 100% completed prior to -initial fuel loading as described in.CO Report No. 219/69-1,; paragraph ^ l II.H. It should be noted that the tests reviewed by Compliance were. l 7 also reviewed and approved by the appropriate GE.and,JC personnal. l l With the exception of two tests, all the ' preoperational tests I required to be 100% complete were found to be acceptable. The two I exceptions were test A-1, Drywell and Absorption-System L'eak Rate Measurement, and test C-14, Standby. Gas Treatment System and Reactor-Building Leak Rate Test. l t

  • Memorandum, O'Reilly to Senior Reactor Inspectors, BWR Main Steam.

Flow Restrictor Nozzles - Possible Design Problem, dated April'1, 1969. t i i t + - ~< ,n,.. ~ -.

i ..b B. JC Review and Approval E As described in CO Report No. 219/69-1, paragraph III.B., i JC management was reluctant to perform the normal review and approval ( function of the preoperational test program. However, based on { continuning discussions with CO, JC initiated a review and approval i program which included direct participation by members of the Plant f Operations Review Committee and specialists from the General Public Utilities' technical support staff. These individuals reviewed all preoperational test results and provided their recommendations to Mr. McCluskey. Mr. McCluskey performed the sign-off approval of all 100% complete preoperational tests except Nos. A-1 and C-14 within the context of the following statement: " Approval indicates that equipment met requirements as described in FDSAR, Amendments or license and Technical Specifications. In accordance with the plant contract, final acceptance of equipment and systems will not occur until after the plant net electrical output performance test." C. Preoperational Test heview Preoperational test results were reviewed during the inspec-tion visits that were made dn February 9 - 12, 16 - 19, 28, March 12 - 14, 19, 20, and 24 - 27, 1969. Prior to the Compliance review, the preoperational test procedures had been revlewed and approved by GE and subsequently informally reviewed by JC and CO. The completed test results were also reviewed and approved by GE prior to the CO review. JC and CO reviewed the test results as they became available following the completion of specific tests. It .j should be noted that JC personnel participated in the test program under the direction of GE personnel. For a number of specific tests, such as the leak rate testing, JC also processed the raw data. D. Preoperational Tests Witnessed The drywell and absorption system leak rate testing and one section of the emergency diesel generator tests were witnessed during the inspections conducted on February 10, 16, 17, March 24, 25 and 26, 1969.

] 4 m..,-. ~~' . i.g! p ,7 , e g. ' 4 s - ) W '4 16'-- ,7 i 3 5 1. Emergency Diesel Generators (D-8) For this ~ test, operation of the emergency. diesei-generator j was initiated by.a. failure of the 4160-volt power source. i' and a simultaneous : indication of a loss of coolant:- L if accident. Operation of the following equipment was d' noted to occur within the time-span indicated and within j Q the time specified in the procedure: 4 g-Component Actuated Tit 6e; Seconds Emergency lighting 0 1 Isolation valve motors 0 i Instruments, controls, small motors O' Reactor building closed cooling water pumps O ~ core spray pumps 0-Standby gas treatment system O' Core spray booster pumps 5 - 10 control rod drive pumps 10,- 15 i containment spray pumps 45 - 50 l Service water pumps'and valves: 120 .f Emergency service' water pumps 120 At the completion of the test, all components were returned to service by manually synchronizing the diesel output with the incoming power. 2. .Drvwell and Absorption System (A-1) q This test'was a leak ' rate determination of the combined' drywell and' absorption system at-pressures of 35 psig and 20 psig. Individual penetration leakage was measured prior to this test. The Chicago Bridge and Iron Company-(CB&I) ' provided both test equipment and manpower for the J -+-.w. me: -es'w.=A -a -+r s eme* %.m er g D

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[ ) [fc ' x ^_ O .. E) g y Te { -- 17._ 4 .) ~! test under contract.to GE. However, Mr. J. Staley,.a 7}' GE Test Engineer, was assigned = as the; responsible jgj iengineer for.this test. Sufficient test equipment. was - q'i . installed to permit leak rate ~ determinations-by.both;the 1 absolute and' reference system methods.. However, the~ 'j ~ A. reference system method was-designated as; primary method 'd for determining the leak rate with the absolute method j! 5: as a backup. The reference system: pressure was greater 1 *, than the drywell and absorption; system. pressure; however, I the reference system' pressure was-monitored.for long P l' periods of' time-(minimum 12 hours) both prior to.and-subsequent to the actual test.to provide assurance that' - the reference system was indeed-leak tight. 4 a. Leak Rate Test Results q The first leak rate--test was conducted during the period of February 15 - 17, 1969, and the following l results were obtained, based on-CO's' evaluation of. -{ the data: 35 psig-leak rate ='1.14 weight %'of the contained air mass per day l 20 psig leak rate'= 0.430 weight % of the contained j air mass per day. j q These test results were determined to be inconclusive j because a water leg was used.to seal the main steam' t isolation valves and various valves were either .j j opened and/or closed during the test period. i + 1 t. At the conclusion of the test, a test instrument { sensititivity check was performed;by bleinding a 'j number of volumes of contained ~ air into the personne1' air lock and monitoring.the response of the various .f instruments. The calculated volume of air that was-released from containment by bleeding through the -l personnel air lock was 2630 cubic feet whilelthe measured volume was 2680 cubic feet, indicating good .i .) sensitivity of the test instrumentation. ~ i l

;p - -.

. l Ra.+, a. 3; m :, ' 1.t: O ' l> x [j * / A second drywell' and absorption system leak rater test was conducted. during. the - period of March 23 L-c 7 I 5 . 25, 1969. For this test, the main steam isolation-4 valves did not have a water seal and valve '1 manipulations were not permitted.. However, prior I to the test, the main steam isolation valves were t exercised a minimum of 6. times.each. The following results were obtained, based on CO's' evaluation of? ' the data: ~ Measured leakage at.35 psig in weight % of-containedairmassperday.[La(35f]: CO value 0.478%' W 208 cubic feet per' hour (cfh)- -GE value 0.476% l v Measuredleakageat20psig{Lm(20]ght%of' in wei contained air mass per day. i CO value 0.469%m l 142 cfh 1 I GE value 0.460%' ? b. Main Steam Line Isolation Valve Leak Rate Testing I,ocal leakage rate determinations for the main steam-isolation valves were, conducted at the conclusion of-. i the test. The local test was performed by placing a I water seal equivalent to'26 psig on the reactor side 1' of the inner. valve. The volume between the two valves was then pressurized and the leak rate'through; the outer' valve determined at a test pressure of-20 -[ . psig. Data associated with the preliminary main; steam line isolation valve testing had a' wide-scatter so' l the valves were cycled and the indicated results,. as shown below, obtained during ratestingt North Valve (A), cfh South Valve (B)', cfh 3.8 28.4 -i _28.4 56.8 'A y my. ---a., ..,w, r.-., __m w . ~,.,

v. ,j ~ . +.. O 3 o 19-- 5

3.,

.9 } c. Comparison, Test Tesults Versus Technical Specifications j-(1)- Technical Specification Limits - All'owable , g operational Leakage Rate .y. l. - Lto(20) 0.75 Lm(20)/Lm(35) = 1 l.' 0.735%/ day - (CO) versus 0.725% (GE) = 225 cfh1 I = s ? (2) Technical Specification Limits -' Allowable Component Leakage 10% Lto(20) = 22.5.cfh Double gasketed seals a Testable penetrations and 30% Lto(20) 67.5 cfh isolation valves Air purge penetrations 50% Lto(20) = ll2.5 cfh and relief valves [ (3) Comparison with Actual Measurcments L i (a) Containment 142 cfh Total Leakage Allowable Leakage 225 cfh j (b) Double Gasketed Seals 1 i I s } (See next page.for list) = -j j-o I ~ 4 =$.4h Ws*4 4."-M. y= 4,F-ar.m.v'e % M 1WWW @' * -W Ag -- 4 SMW -w -.e. r% 's --e. y 4 g.".- O ..) n

f. ~

? Seal Leakage, cfh L. Drywell head d. Drywell head manhole 0 Torus manhole - North 0 'l l.. Torus manhole - South 0 I Tip penetration 1 0 Tip penetration 2 0 Tip penetration 3 0 Tip penetration 4 0 Steam dryer - X-26 'O t Air lock seals 0 Air lock door 0 Total leakage O cfh Allowable leakage 22.5 cfh (c) Testable Penetrctions and Isolation Valves Penetrations and Valves Leakage, ' c fh o Electrical penetrations 0.0 1 Other isolation valves 2.7 4 1 j { Main steam isolation valve-A (North) 3.8 Main steam isolation valve B (South) 56.8* Total leakage 63.3 cfh Allowable leakage

67. 5 c fh
  • Worse case

)

~-.- s.. s/M /. g

7., ~ -

I< + s- ) 1 -..b;... 'i .i .. I 3 .n j I ~ As is. discussed in: paragraph II.A.6.a. of. m this report, the. leakage rate through the 3 A and'B main steam' isolation valves do not-i meet the requirement' described in Amendment. l '? No.-11 (section III, Question 10; page [ III-10-12, paragraph 7.7), which states that j "no valve leakage rate.is to exceed 1.0% of: the allwable containment leakage volume per i r 24 hours." This is demonstrated,as follw s:. i. If the analysis _ is based on the'a11mable .a H operational leakage-rate, then - { i 0.01 x 225 cfh 1 % x L o(20) = t 2.25 cfh = If in the analysis credit is taken for deterioration, then -- l l % x Lto(20) 0.01 x 225 cfh m. .75 .75 l 3.0 cfh i = i 'In either c'ese,.the actual measured. values f exceed that allwed. i i . (d) Air Purae' Penetrations and Relief Valves .h Penetrations &. Valves Leak &qe, cfh V 27-1, -2 -3.7-V 27-3, -4 0.0 l V 26-15, -16 66.3 l V 26-17, -18 20.4 i i V 28-17, -18,. 47 1.1 t Total leakage

91. 5 c fh i

Allw able. leakage. 112.S cfh 9 i i ,:-,~,,.,n.-n- --.---,..-~- 7

[dq w-. ? 4 ,.h O" LO ' J 1 i

,
. ';s P*i

'22 -- 1 o 3 ?s [. - At the time of these tests, butterfly. valve V-28-18 was leaking excessively and could j . g j .not be repaired. The leak. rate of:this l-set.of valves was measured by blanking off ~ d ~ V 28-18'and' measuring the leakage'through U. 1 V 28-17'and -47. Since V-28-18 is in a I series with these two valves s the total: l ^?l[. - leakage-through the set is no worse than-J indicated. E. Preoperational Test Review The following comments and/or background information are provided for some of the preoperational tests that were reviewed. l 1. Hot Intecrated Test l During a test o'f the drywell ventilating system, the l temperature in that portion of the drywell below; the j 0 reactor vessel-to-drywell bellows _ ranged from'90 F to l 120 F. The temperature in the volume < above the; subject. 0 0 bellows reached 180 F. Electrical instruments:are not-located above this bellows however, the wide-range reactor vessel level sensing lines:are located in this j area. This situation may_resultLin false indication. In response to questioning,'Mr. Hess. stated.that the [ electrical components that~are located below the reactor 0 bellows are rated for continuous duty at'135 F and j maximum intermittent duty.at 150o:F.- j t t I I Based on these test results', additional insulation will [ be installed on pipes above the bellows; however, addi-up to confirm the installation of the -insulation 'and ' f tional tests are not planned. Compliance will follow 3 proper operation of the level instrumentation. l 2. Reactor Vessel Components - (A-2) i. The reactor head stud' tensioning operations 7 or a f hydrostatic test was accomplished using three passes i anA two adjustinej passes. ' The total elongat' ion after ' l all passes had been completed varied from 0.053 inches 1 { to 0.061 inches. s i \\ O U' '%r P$**M l "q$ ) W

  • M34
    • i+>"F*e

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    • 'eu

=+**85=* m

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'mh +8 4W*'"" ____4

_, f; v.n w = , ; y_,.2 ]. - ff N Reactor h'ead stud tensioning operations for normal bolt-up was accomplished'using two passes and a single 4 adjusting' pass. The, total elongation after the three f passes'had been completed varied from-0.035 inches to 0.057 inches. y[ t 3. Primary System Expansion (&_-41 Lh; ; 4 During the test, interference between control rod driveT ' thimbles and control rod drive housing supports'was not-checked because the supports were not installed. CO will determine that the adequacy of this installation is confirmed, as is appropriate. Precision measuring devices were not used to verify primary pump-motor plumb specifications. During preliminary discussions, GE indicated that such measure-ments would be made. Mr. Hess rtated that the intent-of previous specifications (built-in cold tilt) was to obtain as.nearly as' vertical alignment as possible.- He also stated - that satisfactory pamp operation during the.- test indicated that the dojective had been met. CO will determine that'the priuary pump-motor plumb specifications-have been verified as is appropriate. 4. Control Rod Drive Hydraulic System (A-5) As described in CO Report No.- 219/69-1, paragraph III.D.2., operating difficulties' were experienced with a number.of. control rod drive units. Following appropriate' corrective c l action and subsequent testing, new problems were experienced. These new problems included double notching, erratic settling' time, and slow wit'hdrawal time. An examination of a disassembled drive, unit l revealed foreign material in the collet, collet piston, 5 and piston seals. A sample of the material was analyzed 4 at GE, San Jose, and indicated that' the material consisted of iron oxide in the form of filings and silica. GE. personnel don't know how the material entered the system. but don't think it was sabotage. Because of the form of the material, they also ruled out corrosion products. E -.e.w. ; m. c.' ., w,. pe, %... w- %d ,,.... w em.- u., A., _p

~~ ' ~ , ~ = yMy y O-1 19, 24 - Vt. ,C[ . During a generalidiscussion,.-Mr.'Hess stated'that the i m - reactor vessel was'very clean prior to the hot-integrated-f"' test. He' then - postulated that the foreign : material-.was . scattered throughout the various primary and auxiliary y . systems ~and then collected in thes reactor vessel'as the J. various systems were operated during the. integrated test.- f The material then entered various drive units' because the "7" ' ; control rod' drive purge' flow could not be operated during this test. During routine operation,the control. rod 1j drive purge' flow continuously flushes all' drive units to prevent foreign material from settling in the drive; units. A' total of 45 control rods were disassembled and inspected. < All 45 contained various concentrations of crud and, as stated by GE, were from a random distribution in the reactor vessel. The 45 control rods that were selected had the poorest operating history (erratic settling and withdrawal times). In response to questioning, Mr. Hess stated that GE could not rule out the possibilityf of crud in'the other drive ~ units. However, he-stated ) that the other drive units-did not exhibit-the erratic f behavior. During further discussion, he stated that-the - GE drive. unit has never -failed to scram at anyc.of the '~ operating faci 1Lities as a result of a buildup of crud. He stated that the drive units will receive a significarit number of tests during the initial fuel loading and subsequent testing program. If additional problems - occur, they will be observed and corrected at that time. Following reassembly, the 45 control' rods,were-scram tested a minimum of five times and two were scram tested 25 l times. 4 Compliance concurs with GE's rational regard 5.ng this l problem and considers the scope of their investigation j to have been adequate. Followup observation will be -{ made by co on the post-licensing performance of the rod drives. 5. Recirculation System and MG Sets (A-7 ) 4 The first phase of a three-phase vibration determination program was performed as part of the preoperational testing program; the other two phases will be performed-en g i 'w w f 4.+ .. h ey's t -mee kmyi.

  • z Efi Ammm44i sesa

.m-a ss.b==+' ib8r rel %# e .DMM---e gr -=d trNg

y i Q.N *_ ~~ .A - ~.~~+~5--. )., J. Jr after initial. fuel loading has -been completed. During i the first phase test, measurements were obtained for -{ operating conditions which simulate those th' t will be a j experienced during actual power operation, except that L the rated pump speed was not included becaus,e of cold d flow pump-motor limitations. The maximum pu,mp speed obtained corresponded to a generator speed of 900 rpm. Vibration of the pump loops were measured one at a time

  • j over the speed range using six sensors.

Two sensors, which were positioned at right angles to each other to measure horizontal motion, were located near the center { of both the suction and discharged lines and on the pump upper flange. Four sensors were attached to the lower inside edge of the flow baffle to monitor horizontal radial motion. Additional sensors were positioned to ) monitor the inside axial surface strain on two control 0 locations. rod guide tubes at two 90 The maximum vibration amplitude observed on the pump loops' was 0.010 inches peak to peak and. occurred at a frequency of 5.5 cycles per second. GE estimates.that [ the maximum stress resulting from this motion would be 2500 psi and that it would occur in the elbgw adjacent to the reactor veusel. GE stated that this,is less than 25% of the limiting value for fatigue failures. Vibrations were also observed.at frequencies of 1.1, 14 i and 25 f* however, the associated amplitudes were quite small and of short duration. q j The maximum vibration amplitudes of the flow baffle were 0.005 to 0.008 inches peak to psak at a predominate-l frequency of 12t GE concluded that these amplitudes { i will result in a stress in the flow baffle of less than g l 1000 psi. j 4 The maximum strain levels observed in the control rod l guide tubes were 15 to 20 micro inches per inch and occurred at 20 to 25.* GE calculated that the allowable l strain for limiting loads in drive housings at the j L measured points is 90 micro inches per inch.

  • cycles per second.

~ _...

wp h f j * '"[ ~" - ~~~'^ ~~ ~ ~ ~ ~ " e ,s-LO. 3 1 i,[ - 26 '-

t_
i. -

.i i-

6. O Liquid Poison System (A--11) l o

@k ' An operational-test of the -liquid poison' system : injection l . capability was' performed at the conclusion of 'the hot .j (( ~ functional test. 'At that time the reactor was hot,:5400 F, j and pressurized, 940'to 985 psig. Injection was-initiated ~ 1 by firing a squib valve and a poison flow rate.ofi33.2' M i gallons'per minute was_ measured'by test tank drawndown-7 with a pump-discharge pressure of.960 - 990'psig.'. The '[ pumps are rated at 30 gpm, 1500 psi each.* f 7. Reactor Refueling and Service Equipment - (A-16) i i The following data is reported for record purposes: a. . Refueling bridge drive speed-

30. feet / minute l

l

15. feet / minute, C

b. Trolley speed I c. Frame mounted hoist, trolley drive l 15 feet / minute speed 1 '.1 d.. Fuel grapple hoists Maximum speed loaded (up) 22 feet / minute. l l . Maximum speed--loaded (down) -: 25 feet / minute.. .l ~ - 8. Reactor Building Closed Cooling Water System (C-8 ) w } As part of the preoperational-test, an investigation'of. the system to survive a loss of one pump was made. 'The [ following information was excerpted from the' test } records: i t The normal heat load flow condition which assumes an average inlet temperature of-1000 F is approximately 4000 gpm. This design flow was'obtained.by establishing 2000 gpm flow rates.in each of two separate paths, combining them and then failing one pump. The operating pump showed no evidence of cavita- - i tion and the pump current increased from 170.to 220 j amperes' (name plate, 235 ' amperes) with a flow of 3750 gym.

  • FDSAR, Volume I, Paragraph VI.4.2.

'i ..n n._.. .. -.. ~. - - -.. ~ - -. - - . - - -, ~ ~.

/ ( s O D ~ 27 - l Throttling of the regenerative heat exchanger was subsequently adjusted for an indicated flow of 4000 gpm and a 71.5 psi differential across the pump. The t .f pump conditions remained satisfactory at a motor current of 220 amperes. 5 I Throttling of the regenerats.ve heat exchanger was - M then adjusted to simulate the extreme runout condition indicated on the vendor pump curve. At this point the ~ flow was 4250 gpm with a 69.0 psi differential across the pump. Pump operation was noted to be satisfactory with a current increase of two amperes to. 222. The cleanup system nonregenerative heat exchanger was sized for a reactor building closed cooling. water flow 0 of 1740 gpm at an inlet tenperature of 75 F. Assuming a 1000 F inlet temperature as a realistic. summer condition, 3800 gpm will be required. This will increase the reactor building closed cooling water, system capacity requirements to 6000 gpm. Based,on these assumptions, an additional phase of the testing was performed. Specifically, with both pumps. operating, the cleanup system nonregenerative heat exchange and shutdown cooling heat exchanger outlet valves were adjusted to establish a 6000 gpm flow. One pump, No.1-2, was then failed and the resulting condition allowed to exist for 25 minutes while personnel monitored for pump vibration and hydraulic noise, which did not occur. Pump current remained at 233 amperes and the 0 F. motor casing temperature remained constant at 110 'I Based on the results of these investigations, GE concluded that the reactor building closed cooling water system can survive a loss of one pump even for the case of a 6000 gpm flow demand. They also concluded that the remaining pump can operate without damage for a period of time sufficient to manually reduce heat load and flow requirements. I v .4sa y V

pls ;fj gj, ~~~~~ Y ' ~ ~ ~ ~ O...c y v.. ~ 28 - ,y j; % ', f: c 9. -Standby Gas' Treatment-and Reactor Building Leak Rate - i:.. - Test (C-14) -1 ,n I.' a. Activated Charcoal Filter Testing i On February 21, 1969,-Mr. John S. West, Barnebey ef 'Cheney company, performed an: efficiency test on: O' the absolute charcoal filtu system; The' test l . as conducted in place using the BC-N77d freon' ll2 w leak test method. A letter, from Mr. West to.GE, indicated that an efficiency of 99.99% was obtained. The Technical Specifications, paragraph 4.5.K.1, specifies that the halogen removal efficiency be not- ~ less than'99 During a discussion with Mr. Hess, the inspector stated that the information presented in.the: letter did not permit sufficient review of this phase. of the test. Mr.: Hess stated that additional informa-tion would be ' obtained. -~ b. Absolute Filter Testing The absolute filter testing was performed bycthe Cambridge Filter Company,using the,dioctly pthalate' (DOP) technique. - Four filter units were tested'in place at an air flow of approximate 1.y 1200 cubic feet per minute. The. test results.wern as follows: Filter Unit Air Flow, cfm Efficiency, % Filter 1 1220 99.952 Filter 2 1200 99.952 Filter 3 1250 99.943 Filter 4 1250 99.930 I Top duct - 1 & 2 in series 99.952 Bottom duct - 3 & 4 in series 99.952- . te7 '?i v ^ m. ,,wr-me.- rw '+

e. me, w

o. .. -wa g yg. 4 %-mm 3.wy. p., u

.( f(

q 7 ym~;

- *' - - - ~ + q.;,; e 1 m. ,') d',: g m m. The Technical' Specifications,. paragraph 4.5L,- g

specifies that the particulate removal efficiency g

be not less than~99% for 5 0.3 micron size particles. i. .c. . Reactor Building Leak Rate Tests 'A number of. tests were performed.to determine the f f' leak tightness of-the reactor. building. Based on-s these test results, the-licensee concluded that the installed components.did not have sufficient capability to satisfy the Technical Specification requirements of maintaining 0.25 inches of water vacuum' under calm l' wind conditions with a filter train flow rate of not more than 1200 cfm.* Another reactor building leak rate test was performeri on March 20,,1969. This test was performed with.10 horsepower. motors on the original fansf*which;resulted in a 1960 cfm flow rate for a single fan and 3400 cfm for two fan operation. An average negative pressure. of 0.20 inches of water was achi.eved.at 1960 cfm and 0.52 inches of water at 3400 cfm. Based on these data, GE concluded that a flow of 2200 cfm will be L required to achieve the negative 0.25 inches of water. pressure. Amendment No.- 52 describes the licensee's proposal; for up-rating the standby gas treatment system. The proposed modification will encompass the absolute and charcoal ' filter systems as well-as the fans. As a result of this -l action, the entire C-14 test will have to be repeated. The test results will~be reviewed by CO when available. t j 10. 125 volt DC System and Supplement, Battery Capacitv Test j (D-12) Battery capacity tests were performed on each battery bank by utilizing a constant 150 ampere discharge source ~. Individual' cell voltage, current and temperature were i taken each hour until the lowest cell voltage reached l l'.75 volts. F

  • Technical Specifications, Paragr.aph.4.5J2.
  • H)riginal motors were 5 hp each.

__;--._,...-.,.a.

- = Q %x a. -^ t.fA ;7,; eg g. =.. J ' /;t} .} 3 j.s 1.c (L ' h . Battery bank B performed very wellland provided a total' ] of 1340 amperes of capacity. A11 cells =in this bank -l were very close'to each other in voltage--and specific l gravity. Battery bank A provided the specified 1200 i ampere hours of capacity. However,~ cell-52, which was: C, 1 within.the manufacturer's specifications, had a lower t f temperature corrected ~ specific gravity than the (,ther_s.- 1 O- 'Mr. Hess stated that cell 52 will be used as the6 pilot 1 fJ cell and require daily and weekly surveillance for cell 4 ,y;v) voltage and specific gravity. 4 j I l F. Compliance with Requirements of Technical Specifications l During the review of the preoperational test program,.this ] writer made a comparison of the test data with the following Technical Specification requirements: l 1. Limiting Safety System Settings, Paragraph 2.3. Limiting Safety System Settings Function Required Actual 1 Neutron Flux Scram 3 APRM (8 channels) 6120% rated neutron flux 8 @ 120% IRM e 15% rated neutron flux Set at 120% of high. range =.less than 12% of rated Neutron Flux, Control Rod Block APRM d 110% rated neutron flux 108% l-Reactor High Pressure Scram 5.1060 psig Four sensors at 1068,; 1068, 1066 and 1066 respectively. The. t

readings shown appear-I to be in excess of that T

allowed because of the necessity to correct for the static head'of. water in the 1:istrument lines. Accordingly, 231" of head is equiv-alent~to 8.3 psig or 1059;7 for the 1068 indicationa.

~ ~ " ~ ' * ^ ~ ~

  • 1

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p ;

y o;; ^ x ~ f 3. y 3d. ', X + fl

1 Limiting Safety System Settinas?

-Function Required Actual .i 7

m.. +

Reactor High . Sensor, Head Correc- :

k. U ;

Pressure, Relief' Valve No. 'psic tionipuia-( . Valves Initiation $ ll25.psig 1 A83 A 1135 =9.b c J 1'A83 B 1139 14.E 'l ]e 1 A83 c 1132 6.8 j 1 A83 D 1136.5 12.2 ] Reactor High 51060 psig with -Four relay combin&tions were; Pressure, Is61ation a time delay checked at-1060 psig and 15-q [ Condenser Initiation 1 15 seconds seconds time delay. i Reactor High 4 @ 1212 psig i 12 psi-4 @ 1212 psig. f Pressure, Safety-4 @ 1221 psig i 12 psi 4 @ 1221'psig-Valve' Initiation 4 @ 1230 psig 12 psi 4 9 1230 psig l 4' @ 1239 psig *12 psi 3 9'1239 psig* Low Pressure Main Steam Line, MSIV .i 850 psig 4 valves @ 859, 860,-860 closure and 861 psigt I Main Steam Line ll Isolation Valve' f.10% valve closure Limit switch set at 90% open { 'i closure, Scram from full open . position + h: J. ' Reactor Low Water jt 11' 5" above top of "O"' instrument indication = . Level, scram active fuel as indi-85 11/16" above'activeifuel, i cated under normal Scram trip established.9 52"' operating conditions which.is equal to 137.11/16. or 11' 5'11/16" ~ Reactor Low-Low 2 7'2" above top of th -Water Level, Main active fuel as indi- ' Steam Line Isola-cated under normal These' two items use same - - 'I f' tion Valve Closure operating conditions sensor but separate relays.** j "O" indication = 85 ' 11/16. : J . Reactor Low-Low 2 7'2" above the top Trip point set'at'5/16";which Water Level, core .of the active fuel, is equal to 86"'or 7'2". Spray Initiation j ~ j i

  • GE will obtain a copy of the certified data.for the outstanding.

valve for review'by compliance at a later date. - **The' acceptability of this arrangement is currently under review by CO:I ; ,u ,r-r ,e--

l, ~.

' =

) [ 32 - h' I 2. ~ Reactivity Control - Control Rod System, Paragraph 3.2', B.3. .L The average of the scram insertion times of all operable '] control rods shall be no greater than: Percent'of Rod Length Required, Seconds Actual, Seconds o P - Inserted Max. Min. 10 0.70 0.36 0.29 50 2.05 0.96 0.82 90 5.00 1.65 1.39 The average of the scram insertion times for the three fastest control. rods of all groups of four control rods in a 2 by 2 array shall be no greater than: Percent of Rod Length Actual, Seconds Inserted Required, Seconds Max. Min. 10 0.74 0.34 0.29 t 50 2.17 0.92 0.82 90 5.30 1.65 1.43 3. Containment System - Functional Test of Valves, Paragraph 4.5.I.2. Closure time, Seconds Item Required Actual Main Steam Line Isolation Valves IE 10 t Valve No. NSO3 A 4.1 Valve No. NSO3 B 3.5 j Valve No. NSO4 A 6.2 Valve No. NSO4 B 5.9 r + y .-.,-----+.e.-- ~... ~-~%,

~

r.

y .e* q_, s ~m I ~ ) J 'd - f 33 - Closure Time, Seconds Item Required Actual b 1 Isolation Condenser Isolation valves i d!60 E Valve No. V 14-30 17.7 Valve No. V 14-31 17.2 Valve No. V 14-32 19.0 S). Valve No. V 14-33 18.2 Valve No. V 14-34 17.5 Valve No. V 14-35 18.9 ) Valve No. V 14-36 18.0 Valve No. V 14-37 19.0 eleanup System Isolation valves n 60 Valve No. V 16-1 13.0 Valve No. V 16-6 12.0 Cleanup Auxiliary Pumps System Isolation 6 60 Valves Valve No. V 16-2

14.0

) Shutdown System Isolation Valve dE 60 Valve No. V 17-1 45.1 Valve No. V 17-2 46.0 i Valve No. V 17-3 43.0 l Valve No. V 17-55 10.0 Valve No. V 17-56 9.0 Valve No. V 17-57 11.0 i IV. Results of Visit - Gilbert This section of the report was prepared by Mr. R. Gilbert. An inspection visit was made on March 12 and 13,1969, to review l in depth the design, installation and calibration of the air sampling and radiation monitor 1ng systems associated with the air ejector off-I - gas and the stack gas waste control systems. Technical assistance was ~ provided the inspector during the visit by Mr. Ormand L. Cortes, Chief Health Physicist, Phillips Petroleum Company, NRTS-Idaho (CO Consultant), and Mr. Larry D. Denton, Inspection Specialist 1 (Health Physicist), CO:HQ, both of whom also contributed to this J report. i d 'f ) A w* 'W.4

  • ' M'g '%.

' 4 t'M. Ws.d Y F -49'*r.444N $N A. 611' M' d M' d4 d y enw.*. m 6-we Mw

r .. ~ -. - --= "A s ) A. General E Mr. Brutschy, Manager, Field Engineering Chemistry (GE-j San Jose), discussed briefly the evaluations made by GE at San Jose in arriving at and justifying the stack gas and off-gas monitoring 0 systems installed in the Oyster Creek plant. Mr. Osborne, Chemist, I Field Engineering Chemistry (GE-San Jose), reviewe<1 the sampling and j monitoring syster.s in use at other GE boiling water reactors and the } various studies (sample collection and analysis) performed at these facilities which demonstrated to them the adequacy of the~se systems in properly monitoring off-gas activity and stack release rates. Both gentlemen stated that the Oyster Creek systems are based on a combination of scientific principles and empirical methods developed from actual applications. B. Stack Monitoring System 1. Stack Flow Makeup Mr. Finfrock stated that the effluents through the stack will be composed of the reactor building exhaust (100,000 cfm design capacity) and the exhaust from the turbine building (90,000 cfm design ' capacity). Mr. Osborne stated that the actual discharge rate may vary between 180,000 cfm and 200,000 cfm but that 190,000 cfm would probably be about the normal rate of discharge. In response to a question, Mr. Finfrock stated that in the event of an accident or abnormal condition which would cause isolation of the reactor building-and activation of the standby gas treatment system the stack j flow rate would be reduced to about 95,000 cfm, with about 4200 cfm from the standby gas treatment system and 90,000 cfm from the turbine building. (As an aside from the main discussion, Mr. Brutschy stated to Mr. Finfrock that JC should be sure that the turbine building exhaust is operating if the reactor building is isolated, to provide a continuous source of dilution in the stack of the off-gas from the standby gas treatment system. Mr. Finfrock assured him it would be.) + t

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-t i 2. Sampling Probe and Sample Delivery Line m 5 The isokinetic sampling probe for the stack gas monitoring system is located 242 feet above ground 7 level (at elevation 265'), at a point where the stack is 15' in diameter. According to Mr. Brutschy, the - n probe and the 1 inch I.D. stainless steel sample line were designed in accordance with ASTM standards. A a drawing of the sampling probe showed it to have four sampling points each centered approximately in equal anular areas within the stack. A diagram of the sampling line shown in a B&R specification manual

  • shows the line enierging from the stack at elevation 264 ', extending out about 4 ' from the stack and re-entering the stack at elevation 254'.

For the remaining 231' to ground level, the line remains within the stack. The purpose of the line extending outside the stack, according to Mr. Osborne, is to enable I-131 and particulate sample collections at close proximity to the probe. A sample holder with a quick-disconnect has been installed in the line at this point. At ground level, the sample line is routed through the stack wall, and around the stack exterior wall parallel with the floor for a distance of about 30' to a panel adjacent to the stack on which the stack gas sampling system controls are mounted. The total sample line routing entails about 275' of travel through six 90 turns. Mr. Brutschy stated that the sample line is purposely routed inside the stack for warmth to reduce condensation from the samples during transport to the monitors. J According to Mr. Osborne and Mr. Finfrock, neither GE nor JC plans to make any determination of the stack velocity profile. The sampling probe position is 10 x stack diameters above the exhaust duct entries. This distance, according to Mr. Osborne, should be sufficient for thorough mixing of radioactive gases. Mr. Osborne

  • Specification S-2299-C, " Piping, Reactor Building - Phase I & II, Main Mechanical Equipment Installation and Miscellaneous Equipment, Jersey Central Power & Light Co., Oyster Creek, N.

J.", dated Augus t, 1966.

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]1: ? 1. stated that he has looked at velocity profiles in other-( - of complete mixing at probe locations in - those plants. plants and the studies have. borne out his predictions k In subsequent discussion, CO stated that the stack f samples must. pass through such a long and tortuous i length of piping that loss of particulate and halogen ([4^ ; activity from the samples through gravity settling, .i impaction, diffusion and condensation would be significant.. If neglected, the samples taken at the ground sampling point would not represent accurately the. activity of halogens and particulates being discharged to the atmosphere. To corre? ate the concentrations measured at the ground level station with the concentration leaving ' t the stack, measurements at the 242 foot height would.be-necessary. Alternately, the filters could be installed at the 242 foot. height. The GE representatives stated that samples from the 242 foot level would be taken and compared with the ground level sampNs~to deterinne a i L correlation. t 3.. Sample Mover and Monitors The samples from the stack are pulled by means of a vacuum pump located at the control panel adjacent to the stack base. The vacuum pump has a variable pumping capacity of up to 3 cfm, but, according to Mr. Osborne, the usual sampling rate probably would be 2 cfm or lower. The vacuum pump is adjusted automatically by a flow i controller which in turn is regulated by a flowmeter ~ i to maintain a constant, pre-set flow through the system regardless of pressure drop across the filters due to i plugging. I At the stack base control panel, the sample is drawn through a heated section of pipe and through a self- ~ sealing, spring-loaded filter head containing a millipore prefilter and a charcoal backup filter. The air then passes sequentially through two separate cup type chambers each of which 'contains a 2" x 2" sodium iodide crystal detector shielded on all sides by 4 inch thick lead bricks. The detector signal output is pre-amplified _7._.

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jggq ) e locally and transmitted through' coaxial cables to:the-i reactor control room recorders and monitor' read-out' panel.- once the sample traverses the system,-it.is-

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-returned to the stack.1 The detector system chambers P and lines 'can' be isolated and purged with dry. nitrogen l C before measuring background.- Mr. Finfrock stated that i 7 the millipore'and charcoal filters will be changed at; "f intervals of'not greater than weekly and more frequently j if higher than normal levels of activity are: detected by. 1 the continuous monitors. The filters will be analyzed { in the laboratory for particulate and iodine' activity; i ~ Mr. Brutschy stated that the stack sampli'ng land monitoring system has been designed primarily for the evaluation.of noble gases; particles having been filtered out by thei absolute filters in the off-gas system (and the standby gas treatment system during abnormal conditions)'and . iodine being present in such small amounts as'to be-completely masked out by the radiolytic-gases. - 4. Calibration of Detectors ~ Mr. Osborne stated that calibration of the stack sample j detectors is accomplished.by the use of liquid samples y spiked with known.1sotopes in pre-determinedL concentra- .j tions. - Using the spiked samples in each detector' system, ] differential curves.of counts per minute of the sample versus amplifier' discriminator' settings are plotted for several different amplifier gains. From these plots,..a family of curves is developed for future reference in checking the linearity of en_ergy response and stability l of the system and for selectirg an-energy bias setting-which will eliminate noise pulses but which is not so 1 high as to eliminate the Xenon-133 energy level (81 Kev), l 4 an important constituent of the noble gas effluents. p J

Assuming that a direct correlation = exists between the count rate ir the spiked liquid samples and the count

'j rate measured in a representative stack sample, the' activity release rate (in uci/sec) through the stack can 'l be calculated based on the. stack flow rate. j -1 -i 4 m -wi* g. r f%2 e n q*+r ma se qWs%,e -*r. ++ qpas, ye ey sesa.-su s.m-w a.a am ew w p - - simp age. yee,aspy..y 4-wi,e we a - ve a se, - -+%-e e Ng : 2..- ,,s

R [? [U mg# h, e - - Lh ', L.l. Ma[.: j-% m,.-.y '*m - 2, w 4: 7;; Mr. Osborne stated that GE is.well, aware of the possible sources of error in the calibrction off a ; gas monitor with a liquid sample but'that'this type'of.preoperational

g calibration'is intended to provide ~a rough approximation

- of the detector respcnse under actual operating conditions p and that the inaccuracies based on the preoperational M;- calibrations'can be overcome'through adjustments of'the .A g instrumentation when the stack releases reach a measurable concentration. Thereafter,' calibration and checks on. 1 the1 accuracy of the stack monitors are based on the analysis of grab samples from'the; air ejector'off-gas. line. .Mr. Osborne stated that:their experience has been that with proper adjustments after startup, this' type of monitoring system is capable of detecting discharges.through" the stac'k as low as 100 uCi/sec when the sta,ck discharge-rate is at its maximum, i'.e.,.when the activity is diluted to the greatest extent possible. Mr. Osbbrne' further stated that the advantages of using liquid' samples for the preoperational calibration of the detector-systems is the relative simplicity of the procedure,.no; non-removable contamination of the system before' opera t tions begin, and an easily controlled and non-hazardous ' means br -checking the response of the detectors and- ~ associated instrumentation over a' wide range.

However, because of the uncertainties inherent.in the liquid' sample; calibration technique 'as compared with the use of a.

-gaseous standard sample, Mr. Osborne stated that GE recommends to customers that. the systems be set to - alarm at one decade below that indicated by the liquid j sample calibration and that adjustments be made in the monitors based on actual samples as soon as' practicable ] after startup. C. Air Ejector Off-Gas Monitorino System .f 1. Samplinq Probe and Sarple Delivery Line The sampling probe on the off-gas holdup line is merely an'open pipe which extends about 10' into the 30 inch end of the holdup line. This is the end where the off-gas lines from the three air ejectors feed into the holdup line. The air ejector off-gas sampling line, 1"'I.D. stainless steel tubing, runs about 100' from the s a u

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L _- ,1 .m, C3 ? i off-gas holdup line to the reactor feedwater. pump room (SE corner of turbine building basement). There the sampled gas is passed through a condenser to extract g a moisture and through a filter and heater before being passed through a vertical section of stainless steel j piping which appeared to be about 5 inches in outside ]: diameter. Mounted immediately adjacent to this pipe segment are two ionization chamber detectors for continuous monitoring of the dry, heated sample stream y as it passed through the pipe segment. One of the detectors is located at the top of the pipe segment and the other at the bottom on the opposite side of the pipe. From this monitoring point, the sample stream is routed back to the off-gas line ahead of the holdup volume. Mr. Brutschy stated that at the point where' the sample is extracted from the off-gas line, the gas is saturated with steam at a temperature of about 130 c. Provisions are also made for taking a grab sample of the off-gas at the detector location. Mr. Osborne stated that a " black box" arrangement developed by GE and furnished to the customer allows for a 14 ml grab sample. The sample bottles are of such size that direct counting can be done with a multichannel analyzer and a 3 x 3 NaI scintillation detector. 1 The nature of the air ejector operations is buch that the off gases are uniform mixtures, according to Mr. Osborne, therefore, there is no need for concern j about isokinetic sampling at the 30 inch end of the holdup line. Also, no consideration has been given to t j establishing a velocity profile within the holdup line. i However, Mr. Osborne stated that sampling experiments l performed by both him and GE cu,stomers have indicated that the profile within the holdup line is bullet shaped l with the leading edge many feet ahead of the trailing edges. Mr. Osborne stated that continuous monitoring pf the air ejector off-gas line is intended primarily for the ~ detection of operational problems and was not designed for absolute measurements of releases for purposes of

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_ 40 _ ( J 4 L pf ct +. 4 y comp.'.iance with. release. limits.--He stated that: i " monicoring of releases for compliance purposes is based f on'oampling'in the stack. g . a 3 e, Mr. Prutschy stated:that, in any event, the off-gas n - monitoring is sufficiently-sensitive and alarms can be - g;l. preset at'such a level that an indication of' abnormally; y yM high or abrupt radical ' increases -in the noble gas - j; activity' levels would signify'that something unusual hadj l occurred requiring investigation and the off-gas system l provides for sufficient holdup and delay to permit i remedial action to be taken to prevent excessive releases.- 2. Predicted Activity in the Off-Gas Mr. Osborne stated that the rate of generation of j fission gases and iodine in the. fuel is a function of j power but that the percentage. contribution'of each fission-i gas and iodine isotope to the total activity in.the coolant and hence the off-gas, depends upon the condition of the fuel. In a recoil mixture, such as is prodyced by gross defects in the fuel cladding (Humboldt Bay cited as an example), there is essentially no delay in release of the fission. gases to the coolant and the contribution of the short-lived isotopes'to the total activity in~the 1 coolant' may be-significant if a sample is analyzed shortly - after it'is obtained. In an equilibrium mixture, such as' l is expected with high : integrity or, fresh', fuel, there' is a very long_or infinite delay in. release-of the fission l]. gases.to the coolant and the longer-lived isotopes. predominate. Between these' extremes is the diffusion mixture, 'in which _ there is an intermediate delay in the releases from the fuel, such as might be expected with pinhole or other slow leaks from the fuel'. 'For each of. these mixtures, the contribution of individual noble gas ? isotopes (22 such isotopes)i.to the total coolant activity can be determined from tables of the' amount of each isotope remaining at given times'after the sample is removed. Mr. Osborne stated that the amount.of iodine isotopes-produced by fission can be calculated based'on power but that the appearance of.significant concentrations ') l 4 $ '.=; i --,,.m m y ~. ~ _..

.: $y_ . _u _ _ a 's ~ in the off-gas is affected by the condition of the fuel, coolant clehnup efficiency and whether the iodine forms g condensibles.. Therefore, correlation of the iodine in g the off-gas with the amount produced by fission and with F the noble gas activity is based on grab sampling in the Ik air ejector off-gas stream, removal of the iodines from j the sample with carbon tetrachloride, and measurement of the iodine activity by gamm spectra analyses. The m {^ GE representatives stated that the iodine concentrations in the off-gas are very low compared with that of the i noble gases and would be completely masked if not measured separately. 3. Off-Gas Flow Meter Calibration Mr. Osborne stated that the off-gas is composed predominantly of radiolytic gas and, to a much lesser extent, air in-leakage to the system. Therefore, the off-gas flow can be determined by combusting the radiolytic gas back to water to determine the rate of radiolytic decomposition of the coolant. Mr. Osborne stated that this calibration cannot be done until the plant is operating, but once the radiolytic gas formation rate is determined, the flow meter can be calibrated by m'easuring the free oxygen in the coolant water which, neglecting air in-leakage, is in a precise ratio to the volume of radiolytic gas discharged per unit time via the off-gas system. 4. Calibration of Detectors l In reference to the calibration of the off-gas monitors, l Mr. Osborne stated that only a rough calibration can be performed before a plant becomes operational. The j inherent characteristics of the monitoring systems of any. reactor facility do not lend themselves to uniform or ] detailed preoperational calibration procedures. The actual radioactive content must be determined and used to perform the final detector calibration. For this reason small calibration sources (e.g. an 80 uCi Co-60 sealed source) are used in preoperational tests to establish a base line from which more refined calibrations are planned and conducted.

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( Or ~l J !' As with the preoperational calibration of the stack monitors, GE recommends that the alarm set pointe be ( a factor of 10 lower than the calibration indicates until the calibration can be verified and the 3 instrumentation adjusted based on laboratory analyses I of grab samples from the off-gas line, j i Mr. Osborne stated that he has developed a' calibration I procedure which is based on empirical data from operating boiling water reactors (GE Vallecitos, Humboldt and Dresden). This procedure is shown to representatives of customers (including JC) who attend a 3 month training course at GE, San Jose. Mr. Osborne discussed this procedure in detail. A synopsis of the procedure follows: The procedure involves the determination of quantities of' fission gases in the air ejector off-gas and the amount of I-131 in the reactor coolant water. Grab samples of both the coolant add off-gas are taken and counted for the activities.of I-131 and Xe-138, Kr-87, Kr-88, Kr-85m, Xe-135 and Xe-133, respectively. The off-gas grab sample is - taken from the off-gas sample line at a point near the off-gas monitor. Two 14 mi samples are taken, one counted immediately using a multichannel analyzer, the second counted 5 hours later. A 3" x 3" NaI crystal is used in the detection system. Working from the concentration of I-131 determined to be'in the coolant water, the steam flow, power level and the off-gas flow rate, a calculation of the noble gas activity appearing in the off-gas can be made. A comparison of the calculated and measured activity of noble gases in the off-gas is made and used to l calibrate the off-gas monitor. i = According to Mn Osborne, this technique, when applied to existing plants, showed calculated and measured j results agreeing to within 20%, an acceptable correlation for calibration purposes. Mr. Osborne stated that a close examination of the measured concentration of each of the six fission gases (in uCi/ml) also provides GE and the customer with information on the condition of the fuel elements. _ - _ _ _. -.~

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~ Mr.'Osborne; stated that additional-counting of the.- s off-gas grab sample together with more computations, 3 sq based'on'a 30-minute-decay time, enables a-det'ermination .1 to be made of the. total concentration'of:22 nobleTgas p,@!! - radionuclides at the stack. 'A comparison of this number-with the actual ~ activity dete'cted by the: stack - g% gas -. monitors ' provides ' thel basis ' for ' operational-1 %M calibration of these monitors. According to-l F Mr. Osborne, exercises at existing plants have show'n. computed and measured activities agree to within 20% - i \\' D. Summary 4 The results of CO's evaluation of the observations made duringl this visit, summarized in paragraph II.A.3 of this report,- were y communicated.to Messrs. McCluskey and Hess by'Mr. Carlson in a j subsequent telecon. y. i V. -Exit Interviews i Exit. interviews were held withl pertinent JC and GE representatives ~9 at the conclusions of.some'of the visits discussed in this' report. . In other instances, exit' interviews were not required due' to' the nature of the visit or becauce the persons involved were?in the. i company of the inspector (s) for:most or all of the visit. -In all y cases, the pertinent items discussed and the significant. comments L made by those interviewed are contained within the. body of the report. qI , e s[ y n o.e- %_I,w ; s +* ,ww m, 4, a u. wn u e 4 L/}}

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