ML20087A120
Text
{{#Wiki_filter:.- -, 9.3 ~. _ _. _,.. ~ .% ~ 1 [ l \\ j -l a: 4-U. S. ATOMIC ENERGY COMMISSION REGION I DIVISION OF COMPLIANCE .g March 12, 1968 k; CO REPORT NO. 219/68-1 ) m
Title:
JERSEY CENTRAL POWER & LIGHT COMPANY 3]' LICENSE NOS. CDPR-15 and SNM-1037 4 l Dates of Visits: November 14-16, 17,.20 and 21, 1967; December 7, 13, 14, 18, 19, 28 and 7 29, 1967; January 9-11, 26 and 30,-1968; and February 5-6, 1968 By r1 Reactor Inspector, CO:I s 7_ . G. Gi bert Radiation Specialist, CO:I M. S. Hildre h, Reactor. Inspector, CO:I .A7 6 F. J. Nolan, Inspection Specialist (Reactors), CO:HQ 4
SUMMARY
Structural work is essentially complete. Most major pieces of ~ equipment have been installed; however, much electrical and in-strumentation work remains to be done. Preoperational testing is underway but has been deferred to some extent pending com-pletion of repairs to the reactor pressure vessel. i j Corrective action on the reactor pressure vessel problems, cracks j in stub tubes and faulty stub tube and in-core instrumentation i field welds, is underway and is scheduled to be completed by l June 1, 1968. Initial core loading is now scheduled to begin } September 1, 1968. A comprehensive and in-depth review of the quality assurance aspects of selected critical systems and other phases of facility construction has been implemented. A special CO task force was utilized in this regard. l i 9500040100 950227 PDR FOIA DEKOK95-36 PDR ~'~
f . da ' l g,a t
- i. 4 j
t + A mceting,was held with the licensee and their contractors on' . January '26, ~ 1968,..to - discuss specific deficiencies and '.other areas 1 + of concern identified to date in. the CO quality asaurance; review.. l ..k Special : emphasis. was placed.on the. fact that the items noted were .' f l. reflective of apparent significant shortcomings 'in the; licensee's l ' overall quality assurance program, and. that further assurances of y a quality product appeared necessary in order. for 'CO to~ arrive ~ at. .i 1 .j al satisfact'ory finding regarding completion of the; facility.. No? ] E firm commitments were made~by the licensee in this matter.. -l ' xg Detailed review of the licensee's programs and procedures for pre-i operational testing, initial core loading and.startup testing has. i continued. A meeting was held with the licensee and their prin - cipal contractor,. GE, on January : 30., 1968,- to discussLthe out - jj standing issues identified to date by this and other reviews. The results of these efforts are discussed-in the report. Major items of equipment and related records were examined in. comparison with descriptions in the. application.. The status of the licensee's program.for the preparation;.of plant operating procedures was reviewed. Draft. copies of the procedures are currently undergoing review by Co. l Defective workmanship, principally faulty welds, has been detected in the prefabricated reactor pressure vessel internals. _Similar l problems have been detected.at another facility
- served by-the
-l same vendor. l -i The licensee is following closely.the problem.with cracks associ-l ated with containment-penetration inserts-to-shell field welds experienced at the Quad. Cities Unit. No.1 and Monticello. facilities ' There is no indication to date that the problem exists at this 3 i facility. j f The possible applicability to this facility of the problem with, poor quality workmanship in shop relding of the control rod drive Sydraulic system penetrations experienced at.Nine Mile Point Unit g. 1 wes reviewed with GE. The installation here has been com-plated and successfully hydro-tested. Further review of this-subject by CO is. planned. Additional observations in the areas of health physics organization, administration, personnel training, equipment, facilities and procedures are discussed in the report.
- Niagara Mohawk Power Corporation's Nine Mile Point Unit No. 1 I
I
.~ .4 %. a -.=/- ^ b 1@f f J.;s A. J
- H.,
4 .s J v 3- } w, 3 lr V ? o V, 1-DETAILS
- I
-l I.. Scope of Visits
- i.
ua 3 sw # .The compliance inspection: team approach to the inspection of the Jersey' Central Power & Light Company's Oyster Creek' Unit No. 1 i, 3 was--continued... Included;has been a'special task' force effort ini j ' the area :of-quality assurance. j e?d .. (-- The~ outstanding issues resulting from the inspection. program: j (f:" ' t. to date were discussed with. the licensee _and their principal con-tractors at meetings. held for that purpose 'at the site on January. 26 and 30, 1968. i This-report discusses briefly the activities of the special-t task force *,. summarizes the results of both the' meetings of January 26 and 30 and of the following visits' to.- the: site: Carlson (Section.II) December 28 and 29, 1967, and i January 9'- 11, 1968-Nolan (Section III) November-20 and 21, and ~ December 13 and.14, _1967' } Hildreth (Section rV) ' November 14 - 16 and December 18..19,11967; and i February 5 -6, 1968 l Gilbert (Section V) November'17, and December 7, 1967 i The principal persons contacted during the above listed visits to the site **were as follows :- J . i, Jersey Central Power & Light Company (JC) e ~1 Mr. George H. Ritter, Vice President 'i l; Mr. Thomas J. McCluskey, Plant Superintendent i ji Mr. Ivan R. Finfrock, Jr., Technical Supervisor 1 Mr. Donald E. Hetrick,, Operations Supervisor Mr. Norman M. Nelson, Maintenance Supervisor [ Mr. Donald E. Kaulback, Radiation Protection Supervisor Mr. Richard D. Doyle, Chemical Supervisor i Mr. Roger W. Sullivan. Assistant Technical' Engineer {
- Results of related visits to be discussed i
in detail in a separate report.
- Attendance at meetings covered in paragraphs II.D. and II.G.
i I
tn .. r- . 1 14~ l ill , ;sl ~ (* ) ~ h .4 - 9_ i ' General Electric Company (GE) } Mr.: Leonard [C. Koke, Resident Manager (Construction) . T Mr. Donald K. Willett, Test and Startup Manager. Mr. Willard C; Royce, Construction Test Director I { Mr. Abel B. Dunning, Construction Supervisor - Mechanical-(Reactor)- l Mr. K. William Hess,. Operations Manager 31 Mr. E. A. Lees, Responsible Engineer - Reactor Vessel Repairs. Mr. Charles Smita, Quality Control Engineer - Piping 1Mr. R. Greene, Test Engineer Burns & Roe, Inc. (B&R) i Mr. Keith E. Clayton, Resident Construction Manager -t Mr. Guido A. La r<i,. Project Engineer (Oradell) Mr. John C. Archer, Lead Supervisor Engineer (Oradell) Mr. Robert Palm, Structural Engineer - Design l Mr. Roy Hettler, Office Engineer Poirier & McLane C6r$1 (P&M) Mr. Tillingham Suffacoole, Project Manager'(Superintendent)~ f II. Results of Visits - Carlson i i A. Construction Status and Schedule Structural work' on the facility is essentially complete with some work remaining to be done on the rad ' waste facility. All major pieces of equipment, with the exception of some in the i i j ' rad-waste facility, have been installed. Visual observations j i indicate that considerable electrical and instrumentation work l remains to be done. Completed and partially. completed systems are t . g in various stages of construction testing. A few of the scheduled. ' l t' preoperational tests have been partially completed. Some tests, t especially those related to the primary-system, have been deferred pending completion of the repairs to - the reactor pressure vessel *. J As of this report,.the licensee is thinking in terms of~ readying the. facility. for initial fuel loading by September 1,1968. This is a change from' the mid-January 1968 date previously reported ** Two principal' factors contributing to the time extension include the
- Discussed in paragraph II.F. of this report.
- CO Report No. 219/67-6, paragraph II.C.
I i .~, r-4 w, Wr ++-av=, -p.o, 4 v-- -,_a ,.---=w-'
- = =.iw e
(RQadL ~ ^' Q%r ~ ' );.[ V
- =
E ({ig.
- (,_)
L) - ~ n >; _5_ LJe qWL ., ' l[ problems.with. the reactor pressure vessel,and-the licensee's A1 dropping of the previously: reported apparent limited ' completion n approach to plant construction. ". ;{.. B., Operating Procedures k The status of the licensee's program for the preparation /f of procedures for station operation was. reviewed by the inspector ~' +r Ar 7jU Procedure preparation, review and approval are being-coordinated'by Mr. Hetrick. He maintains a master set of< records cy <j; for this. purpose. The records list all procedures'and reflect 1 their status. The actual preparation of the procedures -is for the most' part being performed by JC station operating personnel,1with the: assistance of GE in'some areas, including' maintenance, radio-chemistry and reactor engineering (core calculations).--The' general station operating procedures, station systems procedures'and related check lists and procedures for station. abnormalities and emer-gencies.are being written by'the shift operating personnel. Other groupings of procedures such as those, relating to instrumentation, . chemical control, radiation protection
- and reactor engineering are being prepared by personnel in the corresponding portion of the plant organization.
A review of the nester records and discussions with Messrs. Hetrick and McCluskey indicated that.all of the procedures are scheduled to be reviewed by the Plant Operations Review Committee..Also, that all procedures except those in the admin-istrative and radiation protection categories'willLbe reviewed, at' least to some extent, by GE. All procedures will require the final' approval of the Station Superintendent, Mr. McCluskey. According to Mr. McCluskey, the Genera 1' Office Review Board will review critical procedures. He did not elaborate on this-latter point j at this time. 'l' According to Mr. Hetrick, the operating procedures will' be utilized wherever possible' during ethe preoperational. testing, of the plant. Any resultant changes or modifications will require the same review and approval described above. Final review and approval of the procedures will be consummated following success-ful completion of the preoperational testing. p
- Radiation protection procedures discussed in more detail in Section V of this report.
m.. l /t. D "
- o w
, it s...: = ,: 6 h. ) 4 n 7 1 4 All of the operating. procedures are being assembled l into a _ Plant Procedures Manual. The manual breakdown and 1 1; estimated percentage ~ complete in' typed draft f6rm as!of late t ~ January 1968 was as follows:
- k.
Manual Percentage y$ Section Category Complete - !a
- D I p
100 Administrative Proc +dures 75 j i 200 General Station Operating-Procedures 10: 300 Station Systems Procedures 80 400 Station Instrumentation Procedures L30 500 Station Abnormalities and Emergency .10 Procedures .i 600 surveillance Procedures 100 l 4 700 Maintenance Procedures 10 800 Chemical control Procedures 40 900 Radiation. Protection' Procedures 90 ) 1000 Core calculations Procedures 10 i The procedures as described above, including the' program for preparation, review and approval, and the provisions l for proof testing, are in accordance with the FDSAR -Section XII-3, Amendment No. 11 - Section VIII, and the proposed Technical Specifications, j + Draft copies of the operating procedures are being t provided to the inspector and are currently. undergoing review. The results of this review and any' additional observations - relating to the utilization of the procedures during preoperational ! testing will be discussed-in subsequent rept,xts. i i 9 E ,,,,,.,.-n .,,..,p,-- ,-.-,,q-m.w--.,.,,, --_3y
. ~ g v,.;.. ... = -. v ;ff [ I .C) )- ~~ 7- .j _g >I
- + 1
'C. Emergency Plan The licensee's emergency plan, entitled Radiation ] Emergency Plan, is incorporated in.the Radiation Protection j j Procedures portion of the Plant Procedures Manual. A~ draft l .y -copy.of a recent revision of the plan has been provided to the 1 inspector and is currently undergoing review. A preliminary L Q. review indicates that the-plan'is directed primarily to-incidents l -1 that produce high radiati~on alarms within the station. Other incidents,such-as fire and ' hurricane, will be provided for in ,;y a l' section 500 (Station Abnormalities and Emergency Procedures): of the Plant Procedures Manual. The plan was noted to include Lthe following: t 1. Guidance regarding emergency call lists. 2. Action levels. 3. Actions required of station personnel - supervisory and non-supervisory. 4. Provisions.for evaluation of on-site conditions. 5. Provisions.for evaluation of off-site conditions. 6. Guidance regarding protective action for the local population. T The results of the more detailed review of the plan by the inspector, including a determination ac;to its adequacy with respect to the related criteria of the Divi,sion of Compliance *, will be discussed in the next inspection report. l D. Quality Assurance - Meeting with Licensee 1 As was previously reported **, CO undertook to develop and implement a more comprehensive and in-depth review of the l quality assurance aspects of this facility. As a result, a: task i force was formed to Jconduct a special quality assurance inspection-of the following major systems: Recirculation, Steam, Automatic Depressurization, Core Spray, Rod Drive and Containment ***.- The
- Inspection Manual (Reactors), Chapter 0205
- CO Report No. 219/67-6, paragraph II. D. 6.
- Memorandum, J.
- CO Report No. 219/67-6, paragraph II. D. 6.
P. O'Reilly to CO Regional Directors, et al, Jersey Central Power & Light Co., Docket No. 50-219, dated November 16, 1967, Memorandum, G. W. Reinmuth to J. W. Flora, et al, Quality Control Evaluation - Jersey Central Power & Light Co, dated November 21, 1967. .v., ,m ,,--,#..,.,e -ev. +,%.y
-t wh ~'
- ?:
, + ~, - f () ) t, - ./, -composition of-the task force and the results of.the related J. inspection activities will be discussed in detail in the previously referenced separate report. 4 t. Concurrently with.the activities of the special-task .k force, the quality assurance aspects of other phases of facility. y construction are being reviewed, as part of the overall inspection effort, by other members of the CO inspection team assigned lto this' i facility. The results of some of these supplemental efforts-are I discussed -in paragraphs II.E.,II.F., IV.B., and IV.C. of this report. Specific deficiencies and' other areas of. concern ' identified. ; as a result of the inspections performed by both the task force and other members of the CO-inspection team were discussed at a joint meeting, DRL and CO, held with representatives of.the licensee and their contractors at the site on January 26, n1968. The results of this absting are reflected in the meeting minutes.(CO),*a copy. of which is incorporated as' Addendum I.to this report. The results of any additional observations made in the area of quality assurance, including the. followup on the unresolved. ) items identified to date, will be discussed in future reports. E. Concrete and Concrete Reinforcement The facility records show that a total of 87,000 cubic, yards of concrete and 8,600 tons of reinforcing. steel were used in the construction of this facility. I The quality assurance aspects of this phase of construc-tion were reviewed during the December 28 - 29 and January 9 - 11 visits *. Mr. J. M. Varela, Reactor Inspector (Construction), CO:II, i assisted the inspector in that regard during the latter visit **. l The review consisted of an excmination, on a sampling t basis, of pertinent documentation, visual observations of the ) facility structures, and discussions with cognizant licensee and contractor personnel.- The significant results are summarized in i the following paragraphs: i
- Results of earlier reviews discussed in CO Report 219/67-5, paragraph II.A.
- Observations by Mr. Varela included with those of this inspector in this section of the report.
l 2,1 p.- .p., r 1 z-1 .m. g...: ,, ?? % e .i o 3 j l A ~~. 1 ~ A 1. Pertinent-Records-i ][ . Records reviewed included the_following: i . a f a. Work-Specifications. 4 b. Purchase Requisition-(Contract)c-'U. SJ - Testing L Laboratory
- l n
t c. Manufacturer Certifications and Certified Mill' Test Reports:on Materials d. LaboratoryLMaterial Test Reports e. -Concrete Inspection Reports - U. S.; Testing Laboratory i f. Cylinder Test. Reports - U. S. Testing Laboratory g. . Daily Inspection Reports - Inspector'(U.S. Testing: Laboratory) ' Visits ' to ' Batching Plant i h.. Daily Inspection Reports - Burns & Roe Field _j Engineers 1. Procedure.and Welder Qualificat1on Records - I Reinforcement Splices j j. Concrete Delivery Slips I k. Concrete Placement Approval Forms 1. Concrete Placement Records I m. Stripping of Forms Approval' Records i j n. Construction Pictures j i Miscellaneous other records, includingthose o. relating to engineering disposition of problem (discussed in paragraphs E'3 and.E.5.below). l area s l Items of significance noted during the records review are i discussed in paragraphs E. 2 through E. 6. ) 1 l
{ .j mt. wla r mm. < ;.. y 3
- 7..
l ., Gif (} } } i
- e...
O .,y'_, ,?; zgl 'f - 10 ~- S l
- m j
2. Contractors and Responsibilities l m f,.- Drafting of the work. specifications;was done by -Burns. 'i J '& Roe,. project. architect-engineer, based'on design j ~ criteria, supplied by G.E.- 'The. work was performed:in- ~ I -r j several phases by subcontractors to B&R, and under the supervision of B&R and G.E.: z,. y U.S. Testing Laboratory was contracted by B&R to . Provide ~various tests, checks and inspection services- 'l to. substantiate compliance with the specifications j 4 relating to the production and delivery of. concrete. The involvement of JC in,the concrete work was prin-cipally. contractual with GE. 'Ihe work specifications, phases,'and subcontractors are listed below for record l purposes 'I Work Specification Work Phase Subcontractor. S-2299-15 Production & Delivery Eastern Transit P of Concrete ' Mix Co. l S-2299-17 Reactor Building Public Con- .l Foundation structors, Inc. .i S-2299-33 Turbine Building Poirier' '& : ' Foundation 4 & Circ. McLane ~ Corp. - .i Water System 3 Structures l t S-2299-44 Reactor Building - White Con-l Second Phase struction Co. j i { S-2299-45 Superstructure Pcirier & 'i (Reactor.: Building) McLana Corp. - i g + S-2299-58 Reinforced Concrete The Ru,st l Chimney (Design'and Engineering Co. l Build). S-2299-70 Rad-waste Building Poirier & McLane Corp. .L-.,-.-...:. .L.,
k , f :: ; ,w- ' y Q; e 6 ' + . h..)
- ),
~ y (' 70 : j' 11 -- .l 3. Concrete Strength Tests - Period of Low 28' Day. [ Strengths The specifications _ forl the _ production and delivery of a [L concrete, S-2299-15, include the.following pertinent
- g' statements? relating to concrete strength tests
c ~ ' a. "Three cylinders'(will be)~taken from'each 150. j cubic ' yards,.or fraction thereof,. on each day's 7 delivery, whichever-is less,;of each class of-concrete delivered." b. " Test results will be the average of the strengths of the test speciment-----." c. "The. standard age of. test-will be 28 days----." d. "If the average of ~ any five consecutive ' strength tests-----falls below the minimum allowable com-pressive strength at 28 days required for the con-+ crete used, or if more than 20% of the strength-tests have values less than the specified strength, the contractor (B&R) shall have the right to order a change in the cement content or other proportions, of the concrete----." The records show that, with the exception of one period, the requirements in paragraphs 3.a. through 3.c. above were complied with. The excepted period was one in which the test results were consistently below the required 28 day strengths. j The subject period ran from July 27, 1966,;through-October 6, 1966. The affected pours were made under - the contract for the reactor building superstructure, S-2299-45. All of the concrete involved, about 25 pours totalling 2500 cubic yards, was class.3SA (28-day strength - 3000 psi; small aggregater. air entrained). A total of about 75 28-day test cylinders were made, more.than 50% of which were below specifications. The latter ranged from 2% low to 20% low, with an average of 7% low. The affected portions of the reactor build-ing included columns, walls, floor slabs, and portions of the biological shielding exterior to the drywell.
p g 'r.
- A A O A* +
-Jw~ Ad
- 7. i -{f,1 4
h1 .~ s
- y
".12 - l( -The' records-indicate that B&R had the cement. content ~ -y of ' the mix increased twice :Lin an. attempt - to. correct ' f) this' condition but.without success.s Concurrently, }[ .additionalitests.were run on the materials.being:used. { This'resulted in the determination that..the~ problem was , J with the cement *. While in.conformance with ASTM. spec- -q ifications, the cement was found to be unusually _ low- ,. 9 in tricalcium-' aluminate.(3 Cao-A103) and in specific-cr
- surface.
The source of supply was-then chinged**;1the l' n'ew cement was said.to befsubstantially higher.in both areas..The_ subsequent 28 day strengths were within specification requirements ~. B&R conducted-a design review of the structures con. taining the low strength material.- The_ approach used was to treat,the 28 day cylinder strengths as,though - they were the'specified design strengths. The-results showed that there was overstress in only one. member, column B-5 above the first floor. It was determined to be overstressed by 35 A report on this subject *** stated this to be "a condition, not a cause for' concern,. particularly in view of the expected strength after 28-- days." "Ihe inspectors' review of the concrete strength -deficiency problem showed that the problem was adequate-4 f. ly coped with by B&R and ~ that GE and JC were kept in-formed throughout. The inspectors noted from the re-cords that all of the concrete ultimately. exceeded the i specified strengths. y It was noted that the original supplier of cement was subsequently reinstated following correction of the j deficiencies noted earlier. There was no indication I noted of any further problem with the 28 day strengths. 4. Consideration of sulphate Expansion in Concrete in Intake structure Records show that one problem considered by B&R was the possible sulphate expansion of the concrete in the in-
- Supplied by Keystone Cement Co.
- *New Supplier - Allentown Cement Co.
- Letter, D. H. Kregg, Project Mgr, B&R to D. Rees, Project Mgr, JC, dated November 28, 1966.
' U f I ),l .. _, ' ).[ e, ~ i ....n. w-,
- ; y a ',
m./ a ~n w -, n m _- .c. n.. , Q
- .:l
'*t r f (); ) y v. I 'F 13'- g c
- take structure.- It was found - that the 'Barnegat Bay' y
water l contained pulphate'in' the order of 2000 ppm: y T and that in' order to minimize sulphate expansion,i j it' would be necessary to limit the percent' of tri-i calcium aluminate in the cement'to:5% or less.- This;. l 1 meant-that it would be necessary to substitute. Type II ~ q. Portland with Type V Portland; however, the supplier
- q g;-
guaranteed'to supply Type;II with tri-calcium aluminate, not to' exceed 5% with finer; grind.; The records show I that the' design mix resulted'in:the specified strength' at 28 days and with the desired chemical composition. l 5. Faulty Pour Investigation of indicated surface defec';s following. j stripping of forms revealed extensive "honeycombing"- l in at least one instance. The affected. pour.was a 200 ^ cubic yard monolithic pour-that jointly formed a column j and portion of the north wall of.the reactor vessel- -internals' storage pit. These structures had'been11 eft out to allow placement of the reactor vessel in the i drywell. The defective portion was near the base of the pour in column-(D-2), approximately 4'.above the floor at. elevation 95. j During visits in December 1967, a CO inspector visually-l observed the condition before repairs and noted that I there was' considerable loose, segregated concrete that,. when-subsequently cleaned out preparatory to repair, -} extended over the entire width of:the column (4')'over: a maximum vertical distance of 3', and to:a maximum depth of l'. Rebar in the affected area was exposed. l i' .{. U.S. Testing Co., Inc., invedtigated the Condition'for 3 B&R. A report of the results stated thatsreboundz -l i i hammer tests indicated strength deficiencies at the ~ lower portion of the column (defective area),eand I strength equal to or greater than required at the upper section of the column. The. report stated that the con-
- {
crete appeared to have' excessive segregation exposing j crushed stone coated with cement but lackingiany~ fine aggregate or sand content. B&R concluded that the j condition resulted from improper or inadequate vibration during the pour.
- Keystone Cemeht Co.
,,.-,_-r_,, m, ,. - --m e
h 4 ',G g'r n.
- 3. : s :
4 9 '}~.. j. .../ 1 -t. r ~14 -
- N Records'show that the B&R design group evaluatsd the L.
. condition. Their calculations indicated.that the A defective column was satisfactory since it was stressed 2h only to a:small_ percentage ofzcapacity.. The records 9; also noted that the column' design did not take:into' consideration the :added strength provided by the rpool-wall. b 9 117 1968 visit,- thef inspectors During the_ January t I observed that new concrete had been placed to the-original drawing requirement. The"new concrete appeared to be the' same quality as the rest.of the column. In discussing the. concrete defect, Mr. Archer. stated that this was the worst instance of, "honeycombing" experienced on this job.and that the other instances, which were few, were minor in comparison. When asked? what assurance he had that;the specific condition was not indicative of a general condition; he. stated that this instance was detected as the result of inspection procedures normally used, that it can be' assumed that-conditions of this sort.will be evidenced by surface. defects'of.the type noted in this case,.and that no indication of a general ~ condition had been noted. Mr. Willett told the inspector.that as'a result of this finding, GE had conducted hammer tests on other critical portions of-the reactor building. He_said that no defects were'noted. I The inspectors concluded, based.on the above and the results of visual observations made during tours of the facility (discussed'in paragraph II. E.7.), that-3 ~! the engineering disposition.of this problem ~was ade-l} ' quate, and that there was no indication that the con-dition was a general one. 6. Concrete Reinforcement A sampling review of pertinent mill certifications relating to rebar indicated that the results of tests of physical and chemical properties were within spec-ification requirements.
. : r;
- 3.. &
{} ) p t / " t 2 P .The installation specifications required that the bars l .be spliced by the lapping method except for the large L bars (14s and 18s) the splicing of which'would be done
- g by welding or approved mechanical connections.
The f{ techniques actually used were arc welding and Cadweld*. f j. The records relating to required procedure qualifica. fv i tions and welder qualifications, and to the inspection-j of production splices, were incomplete or unavailable to the inspectors at the time of these visits.' Messrs. Archer and Royce indicated that' attempts would be made to collect this information for the next visit. This subject will be reviewed further at that time. 7. Tours of Facility - Visual Observations i The inspectors toured the facility during the January 9-11, 1968 visit specifically to view the. concrete structures. With the exception of the rad-waste R facility,. all major concrete work was complete at the ~ time. It was noted that only about 10% of the interior concrete surface area was " sacked" and, therefore, it 1 could be viewed in its untouched condition. Areas'of-patching, completed or required, were noted to be few and negligible **. The concrete was generally free of ~ shrink cracks. Exceptions included the following: A horizontal cold joint, about 75' long,-at eldvation 78' in the reactor building;' several minor hairline cracks in floors; and cracks at about 10% of the control "V" joints. The cold joint crack 13 at an j interior wall and is mostly superficial, the'same as. l might be expected in successive interrupted' pours of i building columns. None of the observed cracks were considered by the inspectors to be of safety signifi-cance. There was no evidence of moisture seepage, in-cluding at the lowest levels. i
- Erico Products, Inc.
.i
- 0ne exception discussed in paragraph E.5.
) )
,l ld;,p ' a w
- L
~J ; -+ .y - j __. ^,Jfx4 *.; }- 'l s. I' S ' 9 1 - 16' - T l a. No problem areas of any significance perta'ining: J -H' to concrete not_ already discussed _ in ' this-report 'l - 1 were'noted by the inspectorscduring thisLtour. (j# Based on the observations discussed above, the in-spector believes that with the possible' exception'of. 1 r ?gh the area of rebar' splices (to-be reviewed further),. ~ the quality. control program employed by GE an'd 'B&R in the construction of the concrete structures was-I adequate and consistent with that-outlined in' Amendment No. 27*. Also, it1 appears. that: any problems of significance were detected and that proper corrective action was taken where required. During an~ earlier visit, an inspector noted that.. i { repairs were being made to the bottom' surfaces of- ~ several of the large concrete shielding plugs that J 4 go above the= reactor pressure-vessel refueling cavity, at the operating floor level. ' The repairs were to correct defects. experienced during forming of the plugs. The repairs consisted of' replacement. of the defective surface' areas with. concrete grout-as much as 1" thick. The inspector noted that the' 'J repaired concrete appeared to be highly susceptible to damage such that one could postulate loose concrete falling into the refueling cavity..'Unless. detected-and removed,- this : same material might eventually find' its way into the reactor vessel proper, such as during refueling. The problem with the concrete shielding plugs-was-dis- .l cussed with Mr. McCluskey. He stated that he was familiar with the condition but that he did not snare. l the inspector's concern about the possibility of loose ~ concrete falling into the refueling' cavity. He: stated' that even if it did occur, he was confident that -it-1 l would not be allowed to get inside the reactor vessel.
- Except for the referenced amendment, which discusses only the quality. control program and in general terms, no specific references.to concrete and concrete re-inforcement were noted in the FDSAR.
i ~..
~. ~ 'f a if ~ l mdfli Ik i . (), l { Jf. -V o -.17.- s J .1 F. Problems with Reactor Pressure Vessel ^} The subj ect problems ; i. e., cracks in control' rod drive J housing stub tubes ~and faulty stub tube.and in-core instrumentation-1 thimble field welds,1 are being followed closely. by both CO'and DRL. .j ' The pertinent related details have been documented
- and, therefore, -j 4
-will not.be repeated here. As of this writing, GE-JC had. decided" upon the following courses of corrective action: 1. Replacement of the stub tube field welds 2. Contour grinding of the stub tube' shop welds 3. Application of clad overlay to the stub tubes 't 4. Grinding and replacement,as necessary, of the faulty in-core instrumentation thimble field welds. The above~ described work, currently underway, is scheduled l to be compitted by Jube: 1,' 19680 'significant future developments in this subject: area will continue to be documented in -inspection re-( ports and otherwise, as is appropriate. i G. Preoperational Testing, Initial' Fuel Loading and Startup Test Program ** E Meeting with Licensee A meeting was held between representatives of JC, GE and CO, at the site on January 30, 1968, for the purpose of discussing outstanding issues resulting from the CO review of.these subject The results of the meeting are reflected in the meeting areas. minutes (CO), a copy of which is incorporated-as Addendum II-to this report. e The results of any_ additional observations made in these t subject areas, including the followup of the unresolved-items identified to date, will be discussed :in future reports.
- Documentation includes the following:
Amendment No. 29, Status Report - Reactor Vessel Repair Program, 12/4/67;. Memo, Reinmuth to Kornblith, Status of Oyster Creek Vessel Problem, 12/22/67; Letter, Morris to Jersey Central Power & Light Co., Outstanding Issues in Review of Application for Provisional Operating License, 1/9/68; Memo, Price to Commission, Jersey Central Power & Light Co., Oyster Creek Plant, 1/19/68; Inquiry Memo No. 219/68-A, Moseley to ) O'Reilly, 1/24/68. i
- See also Section III of this report.
l 1 ' ~ ' ' l
- ~
- 3
- , _.
c.. dlhY# -... ~ ~~
- p. p w
[.ii$; h . ) 'i 3 d, o. v' :( s. - - 1 ', J: 1 P III. Results of' Visits - Nolan* -._D 1[, A. Preoperational' Test Program - Status l g. iT > the in. si 7 During.the visits specified'inlthisl report, d -spector received copies of outstanding preoperational. test j M procedures. As of this report, all1 test procedures, ex - a capt C-14, Standby Gas Treatment System and'D-3, 230 kv; j Electrical. System, have been received, reviewed by the inspector and appropriate comments provided the applicant. l During the December 13 & 14, 1967 visits,1 both GE and ' J. C. personnel. stated that all preoperational testing would be completed prior'to initial fuel loading.
- In
'l response to questioning, Mr. Willet was unable to' provide' any indication of the-proposed schedule-for _ preoperational' 9 testing and initial fuel loading.. He stated that the control rod stub tube repair-program would delay all test-ing associated with the. reactor vesse1~. He : stated that~ i the repair program was not.sufficiently resolved to provide' a meaningful schedule at that time. -{ i B. Preoperational Test Procedure Approval j A's described in CO. Report No. 219/67-6, paragraph. III. B., Jersey Central personnel do not sign-off on j the preoperational test procedures as an indication that they have reviewed and approved the procedure. In res-l ponse to questioning, Mr. Hetrick stated that Jersey -i ( Central operations personnel do, review all preoperational test procedures and that they have many discussions with [ GE personnel concerning the testing program.. He also stated that they were reluctant to complete the sign-off j j function as a matter of company policy. I i C. Preoperational Test Procedure Review by Compliance l - The following section provides an indication of the status of all preoperational testing procedures; a des-cription of the purpose of the tests that have been re- -l viewed since the las report; where appropriate,l comment on both old and new t est procedures; and an' indication of the adequacy of each test procedure.
- Follow-up to items discussed in this section of the report discussed in paragraph II. G.
W 't W g" e w __;em w ,4,y y ,v.i.v. ,,,.,,,gif,w-s ,w y w-y ,wg.,-3 y,wgy -ew.,- ,9+-+9, y,q gy -yw--<9,-i.g qw -)qg9y, r-9 9+ 9 - weep e 4 w-g ~ g g y' -
3(+. m....v. . ~ y ~. '.. / 3j, - ay' o y w. h j{ - 191-e3 -- mt 1. (A-1) Drywell and Absorption System This procedure was described _in CO Report.No. G. 219/67-6, paragraph III.. c.. The applicant will f(( i. perform a design pressure test:on the drywell at a pressure of 62 psig, and another pressure test {q on'the absorption: system at a pressure of'35;psig. m jj, A combined leak rate measurement will also be per- ~ j formed.on.the drywell and: torus at pressures of 35 'psig and 20 psig. This procedure appears' adequate. 2. (A-2) Reactor Vessel Components This procedure was described in CO Report No. 219/67-6, paracraph III. c., and appeared adequate. 3. _(A-3 ) Reactor Vessel Safety, Relief and' Isolation Valves-This procedure was described in CO Report No. 219/67-6l paragraph III. c. Additional information associated with the safety valve testing program -is that GE. has ordered equipment for bench testing the safety valves on site. Mr. Hess stated that they are also reviewing with the valve manufacturer the manu-e facturer's recommendations concerning bench testing. The preoperational' testing program for_ the main steam isolation valves consists of a functional check of-j' the various controls, interlocks, etc. at ambient conditions.- In response to. questioning, Messrs. j Willet and Hess stated that hot' functional-testing L of the isolation valves will be performed as part T of the startup testing program. - The proposed test, h currently under review by DRL, is;describeduin h Amendment No. 11. 4. _( A-4 ) Primary System Expansion l This procedure was described in CO Report No. 219/67-6, ~ paragraph III. c. The proposed test appears adequate. Based on discussions with the inspector, a coordinat-ing checklist will be prepared for tha tests that will be performed simultaneously. 1
'*h ~ ' ~' jf . + +
- :' L '..
- u. ; -
n hh k,). ~ n L:c y c L _ 20 _ w. d' ^ 5. (A-5) ' control' Rod Drive ~ Hydraulic System The following was the status of review of this I j J -procedure prior to the meeting with the licensee c t., on January 30, 1968 (Ses paragraph II. G.). < i. (( ^ This procedure was discussed in'.CO' Report No. 219/67-6, paragraph III. C.. However,'the follow-ing outstanding problems which include both pre andi_ j post-fuel loading testing of the control rods, require i further resolution. Based on discussions with the inspector, the applicant has expanded the pre-fuel loading, ambient temperature, control' rod testing program from 3 scram operations on all 137 control rods to 5 scram operations on all plus 25 scram operations on the fastest and slowest control rods. In addition, each drive unit will be stroked.8. times. For this phase of the testing program, the appli - cant-wl.11 install two blade guides for'each control i rod. A total of 136 out of 137 drives will also.be~ ~ simultaneously scrammed from the scram discharge volume high water level trip signal once. During 1 a' subsequent discussion, the applicant stated that-they did not intend to perform any pre-loading control rod testing at' elevated temperature condi-tions. Their position was that they did not have 4 the capability of maintaining the primary system l at operating conditions with the control rod drive l-system operating as a full or partial system. They F also stated that testing at reduced' temperature l (250-300 F) would not be meaningful. 0 F h e post-fuel loading control rod testing program was also discussed. In this phase, the blada guides are replaced by fuel elements which -then serve as guides for the control rod blade.- Following the [ assembly of each control cell, which consists of four~ fuel elements and a control rod, the control rod will be stroked to the full-out positionftwice, lowered full-in-once and scrammed once under ambient ~onditions. The-test procedure only described the .troking operations while the scram test resulted from discussions with the inspector. During this ._-.___..__-__________________-_-_-__________-._______m__
m O ~ jfe:p '. h- : ~ = -~ 5 u ,,*7
- ; ;(
? .) \\ -w-1 ,,j> 1 l 9 phase of the program' only one.' control ' rod at a. time j 3 J may be' raised from the fully inserted positiont.- The j [ ' applicant does not propose to.do any control rod I testing at operating temperature and pressure until -l [ the low power testing program has been completed. ] j-. y - At the later date,. they propose to rely on nuclear. j heat to maintain the system at operating conditions.- i u ,~j As-part of the rise to power operation, personnel ~ will monitor control rod withdrawal operations during .I the programmed withdrawa1 aequence from a11 rods in j m b position to approximately'25% of rated power. They ] do not propose to do any scram testing of the rods '~ at elevated' temperatures. They defended their position by stating that nuclear power was required-- to maintain operating conditions and that a control rod testing program would violate the minimum crit'ical~ i heat flux ratio. (MOHFR) requirements if scram-or out-l of-sequence operations were performed.. ll During further discussions, the inspector advised f the applicant that the proposed pre and post-fuel j loading testing program did not appear adequate to 'i demonstrate the satisfactory operation of1the control j rod drive' system. The inspector advised them.to re-sj view the possibility of performing hot functional testing of the. control rods under both pre and pas t-fuel loading conditions. This subject will be re - l viewed during future discussions with the applicant. l 6.
- ( A-6 )
Control' Rod Drive i This procedure was discussed in CO Report No. l '4 219/67-6, paragraph III. C. The procedure appears. -l adequate. A, t j' ) 7. (A-7) Recirculating System and MG Sets This procedure was discussed in CO Report No. 219/67-6, paragraph III.C., and found adequate at that time. 1 ? f
.~z , 4J.[;p, a 1 ~ l , :s s s. [i [) M
- j 4
[? 4 22 - l f 3 -8. (A-8) Emergency Condenser System .i ~! .his procedure was-discussed.in CO Report No. a 219/67-6, paragraph III. C. and found-adequate y at that time. As indicated;in that r'eport, the ] preoperational' test will~ demonstrate that the - .jj various valves and system controls will function- -l ~ under simulated signals. W e hot functional: test-~ l ing has been deferred until power operation has j been' initiated. S 9. (A-9) Reactor Shutdown Cooling System i This procedure was discussed in CO Report-No. 219/67-6, paragraph III. C. Based on a further l T review the procedure appears adequate. ( 10' .LA-10) Reactor Cleanup System i his-procedure was reviewed.in CO Report No. i 219/67-6, paragraph III. C. and found adequate at that time for the ambient temperature, non-nuclear operation.- W e test program will continue during nuclear operation to follow corrosion product and fission product buildup and removal. 11. (A-11) Liquid-Poison System t ^ M is procedure was reviewed in CO Report No. 219/67-6, paragraph III. C. W e applicant is.up-dating this procedure'to better demonstrate the. technical. specification test of the pump by-pass-l system. We appli, cant is also reviewing-the in-j spector's comments concerning.the-need for~ demon-l [ strating the system's capability for injecting .i j water into the reactor-vessel at-the completion j of the primary system expansion test when the system ) is at full temperature.and pressure. Messrs. Hess, .l Hetrick and Willet stated that~such-a test was' ) unnecessary because the system utilized positive. j displacement pumps. nis subject will be reviewed j during future visits. i 5 w -e-%y ,m,. .w e et--"- + - - m.- e eiv= e et.e-- w me-
i
- a..
.z-. n. l 12. _(A-12) Core spray system i l, l This procedure was reviewed in CO Report No.-219/67-6, paragraph III. C. Additional .4 information obtained since that time indicates j that new nozzles that were developed as a .-a result of the San Jose testing program will be supplied. During the test of that system, '[ that has a rated pressure of 185 pai, the t total flow will be monitored and photographs 'taken of the flow distribution pattern. These photographs will be compared with similar photographs that were taken during i the San Jose testing program that utilized i the new nor s. The test procedure also specifies tes,ing under emergency power con-ditions. The procedure appears adequate. 13. (A-13) Containment spray system This procedure was discussed in CO Report No. 219/67-6, paragraph III. C. During subsequent discussions Mr. Hetrick stated that DRL and the ACRS did not give any credit for the system and in fact required more advanced systems to prevent melt down. Because of this position he stated that the system was'not needed and hence did not need to be tested. The inspector stated that Jersey Central. ] should amend the application and delete this i system if it wasn't needed. The inspector also stated that the system would have to be tested if it is not deleted from the applica-g tion. The inspector also stated that the-l procedure should require a quantitative test to demonstrate distribution and that a representative number of nozzles should be monitored in the water test stand. This j procedure will be reviewed during future visits.
,777 ~ ~4 --n e x-n - -_u;-. 's
- 7. s e
(..g -(*}) '). ..u-7 14. _(A-14) Containment Inerting 4 .f This procedure was discussed in CO Report No. 219/67-6, paragraph-III. C., and appeared } adequate.
- 5 15.
.(A-15) Fuel Pool Cooling System } y This procedure was discussed in CO Report -f No. 219/67-6, paragraph III. C., and appeared ' t adequate. r 16. (A-16) Reactor Refuelf.ng and Servicing Bquipment 5 This procedure was discussed in CO Report No. 219/67-6, paragraph III.-C.,and appeared adequate. 17. .(A-17) Reactor Head Cooling System This procedure was discussed in CO Report No. 219/67-6, paragraph III. C., and appeared adequate. l 18. (B-1) Turbine Oil System The purpose of the test of the turbine oil system is to determine the correct and sequential operation of the system. This test will be phased into the preoperational test B-2. Collectively, these tests will prove the relia-bility and capability of the system to function as intended by the designer. The' test appears to be sufficiently detailed and adequate. 19. (B-2) Turbine Control and Turbine Instrumentation i f ,f This procedure was discussed-in CO Report No. i 219/67-6, paragraph III. C.,and appears adequate. 20. (B-3) Steam Cycle The purpose of this test is to prove ' the relia-bility and capability of this system to function, as intended by the designer. The tests will be performed without the steam cycle in actual service.
r- \\ : e$%.r_
- .s
-z ~ .4 . _, a -l r - .d. *?.. L h. ) ] .I ..~w4 n,10 251-- I 4 4J i 1 -in 'The test appears,to be sufficiently detailed'and_ i; l; adequate to demonstrato the simulated: operation-l -' i' of the various. instruments, controls,' interlocks,. j etc., ; rior to generating. steam.. ; However, the i j system will be functionally tested during the. nuclear.- '; phase of the operation. During a discussion the inspector stated that the turbine by-pass valves j J, should be more extensively tested in lightfof the j 1 by-pass valve mal-operation that occurred at other' e I facilities. Messrs. Willet and Hess stated lthat j such testing would be performed. 21. (B-4) Generator Cooling j The purpose of this test is to. prove the reliability and.the capability of this system to-function as intended by the designer, and the ability.of the system to enter into continuous service.- The test-appears to be adequate; j (i 22. (B-5) Main and Spare Exciters .i The purpose of this test is to. determine the correct .and sequential = operation of the exciter switchgear-and protective devices and then followed by a' dynamic: test in connection with preoperational test No.-B-2. j Collectively,'these tests are to determine thei i reliability and capability of the exciter' system to. function as. intended by the designers'; and, the .) ability of the system to enter..into continuous ] service with the electrical ~ generating equipment. s f f 23. Jc-1) Condenser and Auxiliaries system - LI Off Gas System j l l ll The purpose of this test.is to prove the reliability; j and the capability of the condenser system and the - off-gas system to function as intended by the ' de-i ~ signers; and the ability of the equipment to' enter; ~! into continuous service with other elements, that j comprise the complete generating station. ~ 4 w, e
I-l^ - yM' h% _,g,,,gy g, _M- _ge n,g my-gg l - O p i-$ 4 e;i;: 4. -W 1.. gf dl ' ** ~*- e }. f
- P L.L, ;
[l ^26 - N M j 4 The' test' appears'to be adequate..toidemonstra$e T - the reliability of the condenser system..-However, the procedure for the.off-gas phase of the7 test ~ does not provide.~ for any testing of the. filter; L: installation. During a discussion the applicant- ] agreed to perform both DOP. and' freon test on sthe { a absolute and charcoal filter systems:in order that d, m. c anLefficiency determination can be made for.'each I a system. q. i 24. (C-2) Condensate and Feedwater Systems l 1 The purpose of this test is to demonstrate the reliability and capability of these' systems-to-j function as intended by the designer and to enter i into integrated operation with other elements, l that together comprise the complete generating 1 . station. The test procedure appears to be adequate., l ~ A coordinating check list'will be provided for the 1 tests that will be performed simultaneously. [j 25. (C-3) Circulating Wa'ter System i
- )
This procedure was discussed.in CO Report No. -l 219/67-6, paragraph III. C. and appears adequate. ] ~ 3 26. (C-4) Service Water System j This procedure was discussed in CO Report No. l 219/67-6, paragraph III. C. and appears adequate. + i 27. (C-5) Condensate Domineralization 1 System i 'l i j The purpose of this test is-to demonstrate the t capability of this system to maintain the purity J j-of the reactor feedwater and to determine quality, l capacity and reliability.of the effluent. 'Rais l test procedure appears to be-adequate. 28. C-6 Makeup and Domestic Water' Systems and C-7 Turbine Building closed Cooling Water System These procedures were discussed in CO Report No. i 219/67-6, paragraph III. C. and appear to be adequate. l f ...mme n. -m + - ._r -,.,,,y ys ..,%3, e.,.a ,,w-w +- '-
(
- z
[ W R y.7.jp/g n ^.
- v. m, m f
.- 27 W -g j t -p., I '2 9. (C-8) Reactor-Building Closed Cooling Water System 4 Thisiprocedure was discussedjin CO Report No.
- ):
219/67-6,. paragraph III. C. !At that time the II procedure was~ determined to.be. adequate-for the' J [^f system described in the FDSAR.- "Information. h obtained at various ACRS and DRL meetings indicates + . gj y that the basic system may.be. modified to provide' c ? redundant 1 components which will require'a modified' procedure..This system will be-reviewed during
- i future visits.-
~ ~ 30. (C-9) Ins'trument and' Service ~ Air,. (C-10) Fire Protection System, and (C-11) Plant Heating Boiler-1 Tliese procedures were described in CO Report No. 219/67-6,. paragraph III. C. and appear to be adequate. i i 31. (C-12) Heating, Ventilating and Air Conditioning [ System-The. purpose of this test is to demonstrate the ability-i of each system'to perform the design function of: i 'a. Minimizing the spread' of. airborne radioactive - I contamination from controlled to uncontrolled i areas, and to provide safe disposal of such airborne contaminants. l ? b. Protecting equipment-and personnel from temperature'
- extremes, j
The test will demonstrate that the various controllers, interlocks and operating. components:do function and j that the' ventilation-system distribution to the various i compartments and rooms is within the limits specified.. l l However, the procedure does not.specify limits for. l acceptance of absolute and charcoal filter systems. In fact, 'this phase of the test specifies that -the f Halogen and DOP detectors will'be used to st.un the installations rather than to' obtain data for an l efficiency determination. i ) .l =
~,; . : a.. -. w
- l,. (.
_ f L*. .. h }- } 28 - i f, l s f During subsequent discussions, the' applicant-stated j that'the test would-result-in an efficiency. deter-l mination of the absolute and charcoal' filter in-stallations. 32. (C-13 ). Drywell cooling System .:p ~ l The purpose of the drywell cooling l system test'is j 1 to prove the' reliability and capability of this system to function as intended by the designer; and the ability to enter into continuous service with the other elements that comprise the complete generating station. 6 The systems which,'when integrated, constitute the l following: a. Drywell l l b. Nitrogen Makeup c. nnergency Exhaust System d. Station Exhaust System j r e. Drywell cooling System - automatic controls f. Equipment Hatch and Personnel Lock { In general the procedure appears adequate. -However, ) + i it does not specify any testing of the emergency exhaust system under emergency power supply. This subject will be reviewed and ' discussed in future j
- reports, j
t s [ 33. _(C-14) Standby Gas Treatment System ~ This was under preparation at the time of the last. visit. The test procedure will provide'information on the testing associated with the secondary contain-ment. This system also incorporates absolute and charcoal. filter installations that will be. tested by DOP:and Freon techniques. This test procedure will be reviewed and discussed in a subsequent report. w.a ew. m -m-- a + em 6 -P w T
...=-- e; t _ __ _ _ y! --s 4 K h+ ( (). }; i ' h,
- 29 -
i s 1 34. (C-15) l i No test-procedure assigned.to.this number. ' ?
- f.
- {
35.. (C-16) Liquid Radwaste System and-b{. ; (C-17)- Solid Radwaste~ System j s I ' W ese procedures were reviewed in CO' Report No.. l 219/67-6, paragraph III. C., and appear to be adequate.. I 36. (D-1) Electrical Grounding System ~ h is test procedure was discussed in CO Report i No. 219/67-6, paragraph III. C, and appears adequate'. 37. (D-2) 125 Volt Direct current System ) 21s test procedure was discussed in.CO Report No. 219/67-6,' paragraph III. C. W e applicant is.re-viewing the inspector's cominents concerning the need-j for an adequate operation test of this' system under l accident or emergency conditions. W e inspector l described an adequate test as one that includes a-t . capacity. test of the batteries under design load'- l conditions. Mr. Hetrick stated that the utilities-q are reluctant to test batteries under these con-i ditions. In response.to. questioning,'he modified his statement from "al1 ~ utilities"' to "just Jersey ~ t Central." his subject will be discussed.during subsequent visits. 4 } 1 230 Kv Electric System I _(D-3) 38. f W e inspector has not seen a copy of this r.est. 1 -l-procedure to date. 4 I 39. (D-4) 34.5 KV Electric' System, (D-5) 4160 volt Electric System, and (D-6) 480 Volt Electric System h e'se test procedures were discussed in CO Report No. 219/67-6, paragraph III. C., and appear to;be adequate. 4 ee 4 m 9-.- ei r. e w. m.m.c - ~ 9
~ _ _ _ - - ~ ~ - - - - - m: u 2-- > j,, c n. 93,'. [ {} , l} ^' k] ; < - 40. (D-7) Low voltage,- 208/120 Volt'A.C.-System t. t The purpose-of this test-is to demonstrate lthe2 satis-factory operation of the electrical-equipment that-L -serves the' vital-208/120 voltLalternating current f systems for use in instrumentation, power _and pro-
- taction, j
y ',- 1, -The test procedure appears to be adequate. l (U j i 41. (D-8)' Emergency Diesel Generator y y 4 I The purpose of this test is to determine'the correct i as designed operating sequence and operating capa-bilities of the emergency diesel generators and j system _ reliability in providing the emergency power l requirements of the generating station. l The procedure requires testing under the loss of all j , incoming power and loss of coolant incident. W e test procedure appears adequate to demonstrate the operation of the various interlocks,-startup sequences and ability to pick up load on a programmed time delay basis. i The test procedure revealed that one cranking battery. was provided for the two units and that a stepping i switch would select one unit.at a time for cranking.' In response to questioning, the. applicant reviewed-the' installation'and stated that separate batteries were supplied for each unit. 42. _(D-9) Plant Consnunication System I This procedure has been deleted. A check. list will l be provided to verify that all communication stations -{. are functioning properly. ] 7 43. (D-10) Feedwater Heater Controls i The purpose of this test is to demonstrate the reliability and capability of this system to' function j as intended by the designers and to enter into a integrated operation with other systems, i The test procedure appears to be adequate to demon-strate the installation, operation and controls'of this system.
~ l g, y s._
- .p,
- p-~=;". -~:
.} L.g, l
- x n
- 31.- ? N! l 44. _(D-ll) Feedwater Control T-4- The purpose of.this test.istto demonstrate that the h feedwater control valves and control system,w'ill j }f. perform their function 'of supplying feedwater to the y reactor vessel. ].
- }
During the test each of the feedwater control valves' will be tested' separately by bypassing the1 discharge ^ flowback to the suction.of the condensate pumps.-. Test' signals.will simulate operating conditions'to demon-strate that the control valves and control system satisfies functional design and operational require-ments. i The test procedure appears to be adequate. L 45. _( D-12 ) ~ Reactor Protection System f This procedure was discussed in CO Report No.. 219/67-6, paragraph III. C., an'd appears adequate. 46. (D-13) Neutron Monitoring System This. procedure was discussed in CO Report No. 219/67-6,- paragraph III. C. .As. indicated-in that i ) report, temporary instrumentation was-specified for-the initial fuel loading program and low power' testing i program. During subsequent discussions, Mr.'Willet stkted that the permanent instrument. system would be i installed, checked-out and used for all neutron-monitoring operations.- However, he stated that the miniature detectors would be temporarily replaced j i by large detectors to obtain increased sensitivity [ during the initial loading and low power testing programs. 47. _(D-14) Traveling Instrument ' Probe Calibration System -This procedure was discussed in CO Report No. 219/67-6, i paragraph III. C., and appeared adequate.- l 4 ,u, e %~ ,.-ev..~,.. ye --e
ik eweb -,,,e ,, ; } H... ' de m ^ l ws w. A., + e ._o ,a g .t +
- ).
gy n . i ~ j ' 32 - j j . )[ 48. JD-15)- Process' Radiation Monitoring,. (D-16) Area Radiation Monitoring, 1 4. ID-17) Offeas and Stack-Monitoring, and-l 1 (D-18) Environa Monitorinq j i Review of these test procedures-is discussed'in ^d Section V~of this report. { j 49. (D-19) Rod Worth Minimizer I (.. The purpose of this test is to demonstrate that'the 1 rod ~ worth minimizer is operable and functioning'per j applicable sections of specifications l21A 5537:in:- sequences A and B, _and in the shutdown margin mode.- i During the review of this test procedure,~the in-l spector also reviewed' specification 21A.5537.- Based procedure was adequate. on these reviews the inspector concluded that-the D. Additional.Preoperational Tests During discussions the inspector reviewed theLatatus'of j preoperational test procedures that appeared to be missing i from the proposed testing schedule. The applicant-is re- ~ viewing the status of.the following proposed testing - l 1. Automatic isolation of condenser 2. Variable level neutron flux level trip i. 3. Testing to ensure proper installation of flow re-striction and that the installation meets design: 1 i standards. 1 j E. Startup Test Procedures The following was the status-of' review of this subject prior to the meeting with the licensee on January 30,'1968 (See paragraph II.G. ):
- --.i--.,s
,.r
- .-w..
w-+r
n.~. g jaga J } s.- ,., ; t. : ~ pl m ]) f
- j(-
s' } . 1 : j j 1: m 4 During these visit's the; inspector received'a-copy-of the: ) s t. startup; test procedures..1A preliminary _ review'of these: pro-~ l J-cedures has been made. However, basic 1 format'and testing-philosophy will have to be disussed with the applicant: in 1 j ~ i greater detail, to resolve a number of differences.: -Based.onL i ? the preliminary:reviewc thd fuel loading program.and control' rod testing _ program following1 fuel. loading do_not appear -l [4 adequate. The results of:discut;sions with the applicantTwillJ Q, be reviewad in subsaquent' reports. An index of theistartup g. -[ _ test program is provided-in Addendum III. il In~ response to-questioning, Messrs. Willet and Hess stated that the startup test procedures were reviewed by' the supervisor ) of the GE - San Jose group that was responsibic for that' phase. c of the test program, GE on-site startup group;(operation-oriented) and GE - San Jose Technological Review Committee. ] Mr. Hetrick stated that the procedures ~will be reviewed by the- .JC - On-Site Operations Review Committee. The inspector! _ questioned.when and if the startup test program would:be re-' sviewed by a technically oriented review committee sponsored by 1 the applicant-and did not receive an-appropriate reply. This j subject will be' reviewed with higher management'during, future. cj visits. l .j j F. Test Review by Compliance l At the conclusion of the December visit, the inspector- .) stated that he would specifically witness the following pre-operational tests. He also stated that he may witness. other testing if convenient but all test results would be reviewed l in detail. Testing to be witnessed: 1. Primary Containment Leak Rate Testing, (A-1) ~ s - 2. Relief Valve Testing, (A-3) 3. Primary System Expansion, (A-4) ) i 4. Control Rod Testing, (A-5) l 5. Core Spray (A-12) l i ] -r ~,
.; s i, pas-,e .~ -- g 7.-- Q j x c J .y .- 34 1 t 7 T l 6. Secondary Containment Leak Rate Testing, (C-14). y , 'la ' 7. Loss of all Off-Site Power, (D-8) j .) G. -Integrated Test of all Systems . ;:j 7.;; 9. Integrated Test of Nuclear Instrumentation l 9 7 6 i 'i ? l r I i (continued) i t f
- 3 w:e'j...hi
~ ~ ~ ~ ['3 1 '-C s *- 1 <ft: - l ...i. _3 f: 3 ' =p 3 IV. Results of Visits - Hildreth'
- -7m A. ' Startup ' Source - Receipt and Storage, SNM-1037-U l
) ,,3 f. The' inspector' observed that the encapsulated ~5 curie i j . antimony-beryllium neutron source had.been received at the i site and was stored in'the fenced ~ temporary storage area *. ,.jg Facility records. indicate that there was no physical damage %$[fi to the' shipping container-and that no external ~ spreadable-i 9 7 contamination was detected during.the initial inspection at' l 1[ the site. The receipt and storage of the startup neutron source as [ described above is in accordance'with the requirements of i Special Nuclear Material License No. SNM-1037. l B. Quality' Control - Major Items of Equipment i t By sample observations and review of records,- the inspector' determined that, except as noted, the following major item ~s of l equipment were constructed as described in the application **. Name ' plate data 'was transcribed by the inspector. This data is available in the Region I file. i 1. condensate and Feed Weter Systems E The following items were constructed in accordance with the FDSAR, paragraph XI-2.10, and in compliance with ASME Code Section VIII: a. Low pressure drain coolers (3 each). t b. Low pressure feed water heaters (3 each). [ c. Intermediate pressure feed water heaters (3 each). '{ l I p d. High pressure feed water' heaters (3 each). t-Two flash tanks, also part of the condensate and feed-i water systems, as shown in the FDSAR, Figure XI-2-l',. were constructed in compliance with ASME Code, Section 'l VIII. l The steam jet air ejectors are constructed as described in the FDSAR, paraaraph XI-2.4.
- CO Report No. 219/67-6, paragraph IV. A.
I
- Facility Description and Safety Analysis Report, including amendments.
m -T. w" u. msW' ww ,y-o*
m;+:." ~ ~ ^ ~ ~ ' ~ ' .bR + 1[k* t ' .1 Sp l' f '), i :P - 3 6 ' -- .E ,_a m'[' ^; 37 2. Fuel Storace Pool' Filter,' Demineralizing and ~ j Cooling System d Name plate data indicates that the following items of. Y the fuel storage. pool filter, domineralizing andl cooling j system, comply with the description given in the FDSAR,
- Q '^
paragraph X-3.5., and: in compliance with ASME-Code .A section-VIII. 2 .i a. Fuel pool heat exd a.gers (2 each). F b. Fuel pool filter (1 each). l c. Fuel pool domineralizer (1 each). .i The fuel pool cooling and filter system pumps (2.each) j were constructed.in compliance with the description .j given in the FDSAR, paragraph X-3.5. t 3. Primary containment Spray Cooling System The primary containment spray cooling system pumps j (4 each) were constructed in accordance with the~ FDSAR, { paragraph VI-7. l r 4 The name plates on three of the four heat exchangers j in this system showed " Vertical containment Emergency-l Cooling System" in the purpose'section. These' heat exchangers have national board numbers 1412,'1413,'and -l 1414. The inspector determined that these three. items j were constructed in accordance with the FDSAR,' ~ paragraph VI-7 and in' compliance with ASME Code Section VIII. The. fourth installed heat exchanger has a. name plate listing l. its purpose as " Vertical Reflux Condenser" and indicat-l l ing that the National Board Number is 1491. This same name plate had a code stamp indicating that the vessel 1l~ was built in compliance with ASME Code, Section VIII. { t The inspector reviewed the ASME forms U-l for the four 3 heat exchangers which certified compliance with ASME
- j Code Section VIII.
'the four vessels should have -t National Board Numbers 1415, 1414, 1413, and 1412. No information was available from site records con-carning.a vertical reflux condenser or a vessel with National Board Number 1491. In response to questions, j I I
1 , y,[5 _ .Lw; ~- y.- , j y' ' ;.
~
bf/? 11 'r;. I ~ y// ~ .n ] w 4.f i Mr. Dunning stated, that on receipt'of the heat'ex-l
- t changers;for the primary containment spray cooling
-[ system, Lit was noted 'that ~ one.of the' heat: exchangers l 'j .had been labeled incorrectly.. He~ assured the in-m spector that the fourth heat. exchanger (stamped with. National ~ Board Number.1491) was constructed and tested-9.f .the same as the other three. Mr. Dunning also stated [ that this problem would be corrected but would not commit himself to an estimated date. i I The problem with the fourth vessel, either wrong vessel installed or the installed vessel labeled and stamped incorrectly, was also discussed with Mr. t McCluskey.- He stated that-JC would investigate the'. problem and also assured the inspector that it would be corrected. This subject area will be reviewed further on sub-t sequent visits *. i 4.. Isolation Condenser System The tubes in tne two emergency condensers are con-L structed as described in the FDSAR, paragraph IV-3 4 and Table IV-1-1. The shell' sections of the emergency condenser did not have a code stamp to certify that the shells were constructed.in' compliance with.ASME Code lq Section VIII as required in the FDSAR, Table IV-1-1. It is noted, that the shell section of the emergency condensers is vented; therefore,'they arelnot' considered I to be within the jurisdiction of ASME Code, Section VIII**. ~ This subject will be reviewed further on subsequent l 3 l visits ***. 5. Reactor Building Closed Cooling Water System I The two pumps and two heat exchangers in the reactor building closed cooling water system were constructed-l as described in the FDSAR, paragraph X-3.4a. In addition,'the inspector verified that the heat ex- ~ l changers were constructed in compliance with ASME Code Section VIII. J
- Follow-up discussed in paragraph II. D. and Addendum I.
- ASME Boiler and Pressure Vessel Code, Section VIII,
{ paragraph U. l(d) (4). i
- Follow-up discussed in paragraph II. D and Addendum I i
-l 1
y;;bf, ~%Y- = = ~ ~ 4 4 4 s.,< i -p e ?).L e
- . ?--
38 - I 7,. 6. Reactor Cleanup Domineralizer System W e inspector verified.that the following components [ t - of the reactor cleanup'demineralizer system were-con-l J i structed as described in the FDSAR, paragraph X-21 and' j the applicable'ASME Codes:'. i a y~ ( 2 '. each) ~. l a. Reactor cleanup recirculation pumps. P. g L b. Reactor cleanup auxiliary recirculation pump (1 each).. \\ c. Nonregenerative cleanup domineralizer heat exchangers j (2 each). . i d. Regenerative cleanup domineralizer heat exchangers (3 each). e. Cleanup Domineralizer.(1 each). f. Cleanup. filters (2 each). ) I g. Cleanup' filter sludge receiver (1 each). h. Cleanup recirculation pump surge tank (1 each). i. Precoat pump cooler (1 each). j J. Precoat pump (1 each), i 7. Reactor Shutdown Coolinct System From the name plate data, the inspector determined f that the three pumps and the three tube bundles for j the heat exchangers were constructed as described in [ the FDSAR, paragraph X-22. Code stamp or National j Board number for the heat exchangers shells were not-l with the name plate data. S e ASME Form U-1 to certify 3 that the shells were built in accordance with ASME Code, Section VIII, were not available at the site. Mr. Royce assured the inspector that these forms would be avail-able on subsequent visits *.
- Follow up discussed in paragraph II. D and Addendum I.
h I i 1 -*-me v ~g9 ,-y---r e a, m -r m r*4* n'
~ f.. ; _,. _. * .-/p[.. 7 O J A j? 4 8. Steam Jet Air Ejectors ( { The steam jet air ejectors were constructed as I described by the FDSAR, paragraph XI-24. t d 9. Circulating Water Systems + l The four vertical circulating water pumps and chlorination equipment were determined to have been j constructed as described by the FDSAR, paragraph XI-26. 10. Turbine Building Closed Cooling Water System From the name plate data, the' inspector determined that the pumps and heat exchangers for the turbine building closed cooling water system were constructed as described by the FDSAR, paragraph X-3.4b. 11. control Rod Drive Hydraulic System From the name plate data, the inspector determined that the pumps (2 each), control rod drive water filter (2 each), nitrogen volume tanks (137 each), and the accumulator tanks (137 each) were constructed as des-cribed in the FDSAR, paragraph III-5.2.6. In addition,- it was noted that the above items, except the pumps, i which have no code requirement, were constructed in compliance with ASME Code, Section VIII. 12. Liquid Radioactive Waste Control System The following major equipment items in the liquid radioactive waste control system were determined to be constructed as described in the FDSAR, paragraph IX-3.1: I 1
.c.: ,., ;.. e > g.. . _ - -- -~u . ~,..... ' ?: if. j compliance Vented with to Item Number ASME Code Atmosphere . t 'I [ .[ Centrifuge 1 l I .1 . Concentrated Waste Tank 1 X r j Concentrated Waste Mixer 1 ~I Filter Aid Tank 1-X + Filter Sludge Storage Tank 1 X Floor Drain Collector Tank 1 X Floor Drain Filter 1 X Floor Drain Sample Tank 2 X Laundry Drain Tank 1 X Precoat Tank 1 X Reactor Building Equipment Drain Tank 1 X Spent Resin Tank 1 Turbine Building Equipment Drain Tank 1 X Waste Collection Tank 1 Waste Collection Filter 1 X Waste Concentrator Bottom 1 X I Waste Concentrator Preheater 1 X Waste Concentrate Condenser Cooler 1 X Waste Demineralizer 1 X i
- Not Applicable b
.--a
3-x ;;.L ^~~ 3 ;,;,[ --- '-~' .3 ~ ') r" T q f - 41~- p 4-Compliance Vented . f;.- with to j Item Number ASME Code Atmosphere- .i 1 Waste Sample Tank 2 X -nm Filter Aid Pump'for: + Floor-Drain Filter 1 Waste' Collector Filter 1 Fuel Pool Filter-1 Waste Surge Pump 1 Waste Concentrator Feed Pump 1 Concentrated Waste Puup 1 Filter Sludge Pump 2-Waste Neutralizer Pumps 2 Chemical' Addition Pump 1 .l - i, .g t I
_.y
- Vp.jk Ta Q
- .a
> + 4( y o r,;;... . ?.
- 1
- 42~- - i ', {; M
- i.
,V s ,.a - C.. Reactor-Pressure Vessel Internals l l lh Defective. workmanship has been? detected in the reactor' -j i pressure vessel internals.- The. effected equipment _was pre -- fabricated by the P. F. Avery Co. in their shops :in Billerica, j [ Mas's. The results of the initial communication'with the j m license on this subject area are discussed in Inquiry Memo-- .R randum No. 219/68-A, dated January 24, 1968. j JD f ~ ! During the February 5-6, 1968 visit, the insp'ector re-viewed with Mr. Willett the results of-a site inspection made by Mr. Frank C. Rally, GE Design Engineer for reactor pressure vessel internals and Mr. G. A. Berry, GE Quality _ Control-l Inspector.. The objective of the GE inspection was.to deter-mine the' scope of fabrication defects on the pressure vessel internals. These observations were the basis for the engineer-: i ing recommendations-for corrective action discussed in:-this-report. It is noted that Mr. Berry was responsib'e.for GE quality I l control inspection at the P. F. Avery Company shops-during fabrication of the internals. We following items were discussed with Mr.'Willett: o 1. Shroud H6;d and Separator Assembly. - i a. Fillet Welds' Connecting Standpipe to the Inside Surface of the Shroud Head- -t 4 Mr.. Rally. reported that he coul'd not visually in-t spect for weld defects adequately. He noted that two visual defects (slag inclusions) were dis-coverad at.the Nine Mile Site, therefore, he i 'j recommended that all the welds be visually inspected. A dye penetrant examination was recommended for f fifteen standpipe welds, which from the visual in-spection appear to be the most faulty. The writer questioned Mr. Willett concerning.this approach. HeLstated that this was only atrecommenda-- tion and not the GE position, and' that he could not defend the dye penetrant test of 15 welds as selected. by Mr. Rally. Mr. Willett stated that he would. recommend that all welds be given a dye penetrant examination. 4 ,. ~. mi* w s .s.--,,-w-.. ..---.wy r.-,- u..
3~
- p.,
.s. r s 7 A f. ", *.; ~ ~ ~ ~ " " .i ...t y. - ' j[ } J: 4 l, _ 43 _ 1: Mr.. Rally did determine that a number of welds were - ~ undersize,-less than the. required-1/4". 'Because.of. k grinding,rMr.-Rally reported that the size of the l weld could not necessarily be' determined byfvisual. a inspection, therefore,. he. recommended that all_ welds, c, whose size could not be visually determined,-be ultra-- p sonically _ tested using ' longitudinal wave ; techniques. y j He recommended that welds whose minimum ~ size-is 1/8". j or less as_ indicated by the. ultrasonic. inspection be- } repaired. The writer discussed'the GE specifications l-to P. F. Avery which requires a 1/4". weld. (G.E.. Drawing 706E222). If after the ultrasonic inspection only those wel'ds of 1/8" or-less were repaired, this ] could possibly leave welds greater than 1/8" but less' than the required'1/4". In response to questions- ) along these
- 1ines,: Mr.lWillettistressed..that this was-Mr. Rally's recommendation and not the GE position.
j He stated that GE would not have a position until after l 'more detailed inspections of the shroud head and-separator assembly'are made. b. Fillet Weld connecting Standpipe to outside (Upper) i Surface' of Shroud Head Mr. Willett stated that Mr. Rally had reported that most of the welds.were' undersize on the downhill side of the standpipe-to-shroud head connection, and made a a general statement' that only welds on the outer { periphery were accessible for rework. l-Because of the limited accessibility for rework, Mr. l } Rally recommended that all undersize welds or portior.s '] j thereof on the outer peripheral standpipes and the. i second row inboard of the outer periphery be repaired. ] _. f In addition, he reconnended that all accessible under-size welds on the third row inboard of the outer i periphery be repaired. i The writer informed Mr. Willett that this recommenda-tion left much to be desired, since rows 4 through 7 were not accessible, and row 3 only partially access-ible. v'e* i w e -p.,,._m ---y..,7 g yw- -w--
fk %40--- - ~~ ij 2---- ,f J, Nr
- .'+
N O y 'z-44 f-i f;- }L ' i' 4 'Mr. Willett stated ' that ' the ~ GE procurement section-1 had been directed-to make a recommendation concerning' l all pressure vessel' internals repair. This'recommenda-j( -tion would include anianswer to the question, "Should- ~ the shroud -head and separator assembly be returned to _j y P. F.-Avery Companyjfor' rework or should theLrework l be completed at the J. c. site?" After this recom-' 1 mandation is made, GE will take a-position'on rework. l jf [ i.. l
- l f
c. Upper Brackets Interconnectinq~ Standpipes'-(Just, -[ f below separator-to-standpipe weld) ~ These ' welds were ' reported by Mr. Rally to have. numerous defects and, in_ general ~, poor overall j quality -since some brackets were welded!on one. side only. He noted'that only brackets on the outer i peripheral region were accessible for. inspection a nd/or. rework. Mr. Rally recommended that.the loca-i tion of those brackets, which are fillet welded on. one side'only,'be determined using a boroscope.. l Additional recommendations for all brackets connecting-the outermost row to the adjacent inboard row were: .(1) Dye penetrant test all accessible welds. (2) Grind back to sound metal those welds not.in a condition to be dye checked and those having.in-l dications after the test. (3) Roweld to 1/4" fillet and dye check final' welds' as required. (4). Complete missing welds on those brackets that. are welded on one ' side. In addition, Mr. Rally recommended that all accessible-brackets connecting the second row to the third in-board row be repaired. The writer noted that most of 'l these welds were hot accessible, and also noted that no recommendation was made for' rows-3 through 7. ] .l i h e- ....e-e -- - - -. -... ~. - - -... _,.. .-,.-m,
~ 64 ce - s
- a. ; _. ;
W, w '+J. : =+- 1 y' _y ^ T - 45. W g y; d.,' Burrs-in' Instrumented Steam S rator Standpipes ] Mr.Rallynotedthatburrswe(re'found;ina11in- . 7 .M-v
- strumented steam separator: standpipes g the:Nine-Mile Point Unit..Since the Nine Mile Point and JC-T
shroud' heads. wore fabricated by P. < F. Avery Company, at the same time and under one continct,- he.recom - ~[ mended-that all instrumented standpipes be inspected 3~ 7 for burrs, and that the burrs be removed using hand.: 4 grinding methods:only. 2. Shroud a. Liquid Poison System Sparger 'Mr. Rally reported that the required-weld between the saddle shaped brackets and the. liquid poison .sparger was not' preformed. Mr. Willett stated that this weld would be made, and a dye penetrant test made of the-final weld. The writer noted that-this would meet the requirement specified in GE Drawing 706E231. b. Core Spray Sparger t' [ Mr. Rally :noted during his inspection.of' the pressure vessel internals that'4 nozzles on the' lower core spray-ring'were less than 1/2 inch from the support bracket for the lower and upper core - spray. rings. He recommended that the bracket be } cut in the center and the lower half moved so that the lower nozzle would be-more than.1/2 inch from .j the bracket. The inspector questioned Messrs. j Willett and Dunning to determineLif the 1/2" j clearance had been specified. The specifications j-were not at the. Job site, but Mr. Dunning thought that Mr. Rally had made recent. calculations and determined that 1/2" was required for expansion. Again, this is only a recommendation by,Mr._ Rally; therefore, no work will be done in this area until the site receives detailed instructions from' San Jose.
1 -Q4 1.* : ~ .w gy.7, w ;;.
- g
- 3 ~ ~ .46 - yj e -i ,3 D.' Containment Penetrations Inserts-to-Shell' Field Welds q ,In discussions with Messrs. Willett and Dunning; the in- .L spector determined that the site personnelzwere; informed con-- 3-
- cerning recent problems in'this area *.
Mr.-Dunning stressed < i that the! plate used at JC waslA-212-61T, Grade "B", made:to l ASTM A-300. requirements,-and that"the minimum charpy vee notch 'j Limpact test value was '20. foot-pounds : at 0 F. instead of '13 foot. 0 l. -pounds-on full. size' specimens as permitted by Code. Case 1317. P He noted that this plate'is not made of the'same material'as-c used in plants where the problems have occurred. According ~ 'to Mr. Willett, all penetration insert field welds ~at JC i were made during the summer months,'as described in Amendment i Number'15. Mr. Dunning stated that in addition to the quality control tests described'in Amendment Number 15, all vent: system penetrations and the equipment hatch had been-given a magnetic i particle. test. He. stated that no defects were. located.during i these tests. Mr..Willett does not think that JC has.a problem in this area, but he is following this with other GE plants under con-struction. He stated that he has no plans at present to-con-l duct additional tests. However, if'it develops that the problem has applicability at JC, GE would re-evaluate this-position. y E. Control Rod Drive Hydraulic System Penetrations 2 The' quality control of shop welds, both at the coupling and at the point of penetration,'were discussed-with Messrs. j Willett and Dunning..Both were aware of the reported problem at Niagara Mohawk Power Company.** ~ Mr. Willett stated that the j hydraulic lines h4d been hydrostatically tested at"150% of- .{ design pressure. No leaks were reported during these. tests; l ~! therefore, the GE position was that no additional ~ tests were f } required. The inspector requested reports of quality control tests which were conducted at Chicago Bridge & Iron company shop. These were not available at the site. Mr. Dunning said that p this information would be made available at the site in the near future.
- Quad Cities I and Monticello
- Memorandum, Moseley to O'Reilly, Niagara Mohawk Power Corp.,
Docket No. 50-220, Defective Welds, Control Rod Drive Hydraulic System Penetrations and Reactor Pressure Vessel Internals, j January 23, 1968.
w~ y l % L J ' w ~
- ^
.i %Hf?M. [j % p[ ,, :j - + i L i 4 This area.of concern will be reviewed further on ' W subsequent visits. j iq. k 'V. - Results of Visits - Gilbert l J' i U A '. Health Physics Organization and Administration ~! Q$ [ i n e' inspector discussed with Mr.Kaulback the status of-the j ^ Radiation Prot'ection Procedures (Section 900, Plant Procedures .{ g Manual). We first draft of these procedures has' been published. i and a ' copy provided to OhI. We results of this discussion are covered in paragraph V. D. ? The inspector d'scussed with Mr. Doyle the status of the chemistry and radiocher /.stry procedures (Section 800), and counting i i room procedures (Subsey: ion 909). Mr. Doyle stated that most of the subject procedures Mere in the process 'of being. typed and-printed. He -indicate! these procedures would be available within several weeks and CO:) would receive copies. Some of the: counting. 1aboratory and radioci emistry procedures are being held up until a specific counting equipment arrives, ' and until Mr..Doyle has.'use~- 1 of_ the radiochemistrN 1aboratory to develop. procedural techniques.- { W ese matters will bi reviewed during subsequent inspections.- t j -l During norma ( plant operations, JC plans on having regular - 'l Y-health physics cover l.e only during the day shift. - JC employees j will work a five-day ! 40-hour week. On weekdays, the Radiation Protection Superviso:' Chemistry Supervisor, and at least one-HP + j - technician, one ~ chemf try technician, and one assistant technician, 1 ^ will be on duty. We / end work will be performed by one HP. tech. { L nician, one chemist i technician, and an assistant technician. d i i j With the ex t ption of scheduled walk-around surveys to be made by the shift ff eman, or a shift operator, on back shifts, l E regular health phy-/ cs and radiochemistry work will be done during j the day. If 4 prcriem develops on a back shift, which the foreman j determines h4 can /at handle, Mr. Kaulback or a technician 'will' be called for issist/nce. According to Mr. Kaulback, ' foremen and operato.J will b/ cross-trained in HP activities and capable of handlivj all butjemergency situations *. We inspector indicated j
- Prev / susly discussed in CO Report No. 219/67-6, paragraph V.C.
l 1 H l ) i 4 4 . ~.
C .L __ 1 I O ,.) < to Mr. Kaulback that this approach is acceptable, but only after each foreman and operator has undergone on-the-job training under { the direct supervision of the Radiation Protection Supervisor (RPS). t When the RPS is confident a person can perform backshift HP duties t in an acceptable manner, this fact should be recorded. 1 The cross-training program and qualification of station personnel I in HP functions will be closely reviewed in subsequent inspections. According to Mr. Kaulback, requests for the purchase of radio-active material will be prepared by him and sent to the Technical Supervisor for approval and, if necessary, preparation of applica-tion for license amendment. Receipt, storage, control, shipping and required radiation surveys of radioactive material on site will be under the supervision of the RPS. This general operating method agrees with that specified in Amendment No. 11, pages VII-1-2 and VII-1-3. HJwever, as noted later in paragraph V. D., the Plant Procedures Manual does not discuss these areas in great detail. B. Training Program
- The first basic training (indoctrination) sear".un in radiation safety was held during the period November 13 to November 16, 1967.
Six persons, station helpers, electrical and instrument maintenance personnel, and stores personnel, with no previone knowl ' edge or training in the nuclear field, attended this course. About 27 hours was devoted to this session. It was conducted by Mr. Kaulback and his two health physics technicians. The rudiments of radiation physics and safety were covered. Essentially each day was devoted to two hours of lecture, two to three hours of problem solving review of subjects covered, and preparation for the next session. Attendees were expected to do some off-hour study and review. Tests were given and graded. NT. Kaulback stated that he thought that the course was successful and that attendees received sufficient knowl-edge of the subjects to enable them to attend more advanced and specific futuro training. The second indoctrination course was given s during November 21-27, 1967, and was attended by seven station em-ployees. The records pertaining to the training sessions were re-viewed. They reflected names of attendees and grades of the tests given. No grades below 70 were noted by the inspector. According to Mr. Kaulback, and as confirmed by the inspector, the basic train-ing was developed in accordance with statements in the FDSAR, Vol. I, Section XII. Detailed course content, beyond the basic material in
- Previously discussed in CO Report No. 219/67-6, paragraph V.B. and V.C
g,n -} 4 - o .9 O -f i - 49 1 di 's t p the Radiation Protection Procedures, and additional-training. l T, '. sessions'will be~ reviewed in a subsequent inspection. j 3 C. Equipment and Facilities 1 i According to Mr. Greene, the local readout and ' alarm ' boxes ~ 'M 10 ! for four radiation. detecto'rs associated with the area radiation monitoring system have been' mounted on the-119' level-(operating 9 4 j floor) of the reactor. building. Cable has been pulled,for:the M- ^ four-detectors, but final connections have not been made. l .l During a tour of. the facility' on December 7, J 1967, Mr. t Kaulback pointed out to the inspector the approximate location of ] each of the 30 area radiation monitors. The'four installed readout; j and alarm boxes. at the 119' level-were observed.= The inspector j noted that boxes for detector numbers C-9 and C-10, fuel pool bridge .l low and high range monitors, respectively, were not mounted on the 1 fuel pool bridge as. indicated in the Radiation Protection. Procedures' 'l section of the Station Manual and in Pree' t tional Test' Procedures No. D-16. "The monitors were placed on thI surth wall--of the reactor building operating floor, about 20' from the pool. Mr. Kaulback i stated' that efforts are being made to have GE either move the i monitors to the bridge or to provide an alternative method of monitoring the bridge area. This matter will be reviewed during;the next visit to the site. The tentative locations of the'other 26 monitoring stations appear acceptable and should accomplish the purposes specified in the FDSAR, Vol. I, paragraph VII-6 3.1.- i Locations of the process radiation monitoring and the environs monitoring stations will be reviewed during subsequent site-
- visits, f
The detectors for all radiation monitoring stations-are j on site and stored in the control room. Electronic equipment for j l the monitors is completely installed in the control room panels-l ] and has been bench tested to the extent possible without detectors. No firm date has been established for completion of the radiation l monitoring systems. GE has recently let the contract for~ fabrica-j tion of two cabinets to house the two outside environs monitoringL j systems. F l j f x i . -, ~ - - -
i w.u-O ^) t l - 5G - l i Final plans for calibration of radiation monitoring equip- 'i [ t ment.and filter' testing have not been developed. As discussed in I CO REPORT NO. 219/67-6, paragraphs V. A. and V.E., Mr.'Doyle.is f responsible for calibration' of process radiation monitoring systems, i I, (Mr. Kaulback will have the opportunity to approve these calibration procedures), and testing of various system particulate'and halogen-j filters.. Filter testing will' include.in-place' tests. Mr. Doyle . } } indicated he was considering an in-place test similar to that per-l e formed on the N.S. 3AVANNAH. %is approach.is in accord with state - j ments incorporated in the Application, (FDSAR Vol. I, paragraph i ' -V-2-8 and -9), and Amendment No. 23,- pages 4 - 13 and 4 - 15, relating to the Standby Gas Treatment System. Additional review will be made when final filter testing procedures have been formu-~ { lated and published, especially those relating to filters in the air ejector off-gas line, gas treatment, and the stack gas moni-- { toring systems. Mr. Kaulback will be responsible for calibration of area and environs radiation monitoring systems. I D. Plant Procedures Manual - Radiation Protection Procedures -(Section 900) h e contents o'f the Radiation Protection Procedures portion l of the Plant Procedures Manual was reviewed by the inspector. ' We - l procedures in this section establish what was intended; i.e., set i forth the basic philosophy of the radiation safety program, and j provide the general radiation protection standards and procedures. Section 900 is not a Health Physics Procedures Manual, per se, and [ as such, does not fulfill the needs of providing detailed instruc-tion and specific procedures. W e subsection on various records - with Mr. Kaulback, it appears these subsections will not add any and reports are still not available; however, based on discussions. I appreciable degree of specificity. - i .{ With considerable verbosity, Section 900 discusses'the j. following: area control (definitions, radiation'and contamination ,j limits, and access control); personnel protection and control; respiratory equipment; emergency procedures (hospitalization, major and minor accidents, and the Radiation Emergency Plan); I equipment control; and, transportation of radioactive materiale, j Storage of radioactive materials and radiation monitoring are i discussed rather briefly. ) P-' e .--+.v 4 m w. _.,,y y w ,,.%ig --y g-
- g,-
s'y 9 gow--
~ n. 0 i s' -i. 1 l - 51'- The 'information on hospitalization of contaminated i casualties, Subsection 905.13, and the Radiation Emergency Plan, 3 Subsection 905.14, is the'same as that specified in the Application, } Amendment No. 11, Section VIII 1. i The absence of specific instructions and procedures in 'i j the manual was discussed with :Mr. Kaulback. He stated that the t ~p Procedures Manual would be supplemented with oral, and in some a instances, detailed written internal HP procedures,-.which would j i be furnished to HP technicians
- and other ' station personnel.
Preparation of some procedures would be accomplished after the plant is operating. The following were pointed out to Mr. Kaulback as general areas which should be presented in more detail-l 1. The frequency of and specific areas for making routine i dose rate, air particulate, radiogas, swipe, and grab sample surveys. l t 2. Situations which would require performance of special (general or local) surveys. 3. Surveying and sampling techniques. 4. Sample counting techniques and procedures. 5. Equipment calibration frequencies 'and procedures i (including the fixed radiation monitors). 6 6. Radioactive. waste management systems. i 7. Control of radioactive materials (ordering, receiving, j storing, inventoring, and leak testing of sealed 1 g i sources). i I ~ 8. Laundry operating. procedures. f I 9. Containment entry procedures (including preliminary j surveys). 10. Contamination control and decontamination procedures l (including entry and exit procedures for controlled l areas, donning and removing protective clothing, l l i l
- - - - - - - - - - = - - - - - - ^ ,fg - ~~" p Q'?i [*i e: ,it:l 0 [] _' J Q;Q j [ [, g 1 ' survey of protective clothing and respirators, -] -and respirator filters before and after use). J j 11. Reporting of-loss or theft of radioactive material; q. 12 '... Reporting to AEC and to individuals excessive ex- ,a p W posures to radiation and airborne concentrations.- i l The development'of written HP procedures relatiing to' the p above'will be reviewed during subsequent inspections.. l?[ E. ~ Preoperational Environmental Surveys - r t The basic purpose of the preoperational radiological-surveys of the oyster Creek Station environment,.as stated in the ~ FDSAR, Vol. I, page I 3, is'to provide a base' line from which increases'in radiation can be evaluated and related to plant operations. The six general types of monitoring to be done'were as follows: 1. Atmospheric Radiation 2. Fallout I 3. Domestic Water -i 4. Surface Water 5. Marine Life j 6. Food' Stuff 1 i Additional details on the program' appear'in the FDSAR, jL vol. I, pages 11 e6el to -3. Thirty-one locations (30 off-site-l and one on-site) ware to be used for data collection. Statements ~' indicate the method of monitoring, number of sample stations, sample frequency, and analysis schedule for each type 'of monitoring. .4 4 m- _ _
J. -~._. l o . r The raw data accumulated by JC on the preoperational e environmental surveys were reviewed by the inspector during the h December 7, 1967 visit. The results showed that t he surveys T had been performed as specified in the Application. Monitoring ( of food stuffs (crops) started in September 1966, the other five general types of monitoring were effected during the period March - f June 1966. Isotopes, Inc., Westwood, New Jersey was contracted to ,j perform the analysis of various survey samples. The film packets used for radiogas sampling were provided and analyzed by the Radiation Detection Company, Mountain View, California. For reference purposes, Mr. Kaulback had tabulated the results of various U. S. Public Health Service and U. S. Department of Interior, water pollution control, environmental sampling programs undertaken to evaluate radioactive content of rainwater, surface water, airborne particulates, and milk in the nearby Pennsylvania, New Jersey and New York area. The JC results were noted to be essentially the same as those of the federal agencies. F. Preoperational Test Procedures 1. _(D-15A) Process Radiation Monitoring System - Vent and Effluent Radiation Monitoring Subsystem The stated purpose of this procedure is to demonstrate that all vent and effluent radiation monitors are operable, in calibration, have correct trip settings and perform specified annunciation functions on up-scale or downscale trip. These procedures cover the testing of radiation monitors for the reactor building ventilation ex-haust plenwn, isolation (emergency) condenser vents, i and the containment spray heat exchanger service water effluent. G-M tube detectors are used in each monitoring system. The ten detectors in these monitoring systems will be placed as follows: Two in the ventilation exhaust plenum (one each for vent manifold No. 1 and No. 2), four for the emergency condenser vents (two each for vent No. 1 and No. 2), and four for the containment spray effluent (one on the service water effluent of each containment spray
4 XMLiu1-f.j Q- ' ' :. ~ k r.l Q ?) 7, 7-+ .54 - t s - heat exchanger)..One function of the' ventilation ~ex- . t-haust plenum monitors, in conjunction with the - fuel - '[, pool and-reactor head ~ area. monitors, will be'to activate'
- O the gas treatment system according to' preset conditions.
M h e' test procedure appears to be sufficiently detailed ,j' except for detector calibration. Mr. Doyle: stated that-i he is' developing. calibration procedures to augment the; f'9 GE procedure. Development will;betdone in conjunction with GE representatives. Wis preoperational test i procedure will be reviewed further when the~ details on, detector calibration are made available. 2. (D-15-BJ Process Radiation 'Monitciring - System r Process Liquid Monitoring Subsys g W e stated purpose of this procedure is to demonstrate-that the liquid process monitors are operable,.in - calibration, have correct trip settings and-perform specified annunciation functions on upscale or downscale' trips. W e procedures cover.the testing of_the' liquid radiation monitors for the rad-waste system discharge, reactor building cooling water and reactor building service-water system discharge.' In.each' case a scintillation-detector is located in a shield which surrounds the water containing pipe. We test procedure appears 'to be. satisfactory regarding .i electronic and recorder checks.and should adequately b demonstrate the system. S e details on detector cali-bration were lackingt however, Mr. Doyle has stated,his 4 intention to supplement these procedures with a detailed f procedure which will use a calibrated : liquid radioactive - [ source. M is procedure will be reviewed further'when .+ the' details on detector calibration are made available.; . 3. (D-15-C) Process Radiation Monitoring System - Steam - Line High Radiation Monitoring Subsystem W e stated purpose of this procedure is to demonstrate that all steam line radiation monitors are operable, in-calibration, have correct trip settings, perform specified m e y- +p
- 1m pw
+-y.. e-e w-t 66 g>. 9
y[f[ Q ,a n '. %pa,._:A - ^ ~' s ev 4-l .y ff Q } L i r{ 4 ' 4 z. 3 annunciation functions-on upscale;and downscale trips, ~
- and ' operate scram relays L in Panels.6R and: 7R ' (close y
q> main steam line' iso 16 tion-valves).-- n [ ' Four' channels of. instrumentation are covered by these. procedures. Two each are adjacent to two of the nain y nC ~ steam lines downstream of the outer isolation valves - j . at theLdrywell' penetration. Each instrument provides 'l' j. for continuous gross gansna monitoring of each primary q j' steam line.. Ion chambers are used as radiation detectors. J Although the' test procedure appears sufficiently detailed [ g and adequate to demonstrate the system, it lacks' details regarding-detector calibration. Mr. Doyle plans'on-developing a calibration procedure in conjunction with-GE representatives which he says will probably use a i calibrated liquid source. This test' procedure will be reviewed further when the details on: detector calibration { become available. i 4. (D-16) Area Radiation Monitoring I The stated' purpose is to demonstrate that all area j radiation monitors are operable,-in calibration, have j correct trip settings.and perform specified annunciation j functions on high or' low trip. Thirty gamma sensitive detectors monitor the radiation levels at thirty, locations where personnel might'be l }. working. In addition ~to activating' control panel [ annunciators, some of the monitors are equipped with i U local readouts and audible and visual alarms. Twenty-nine. of the monitors cover one of two ranges, 0.01 - 100 mr/hr-or O.1 --1000 mr/hr. One monitor, ghefuelpoolbridge' -l high range, has a range of 10 to 10 mr/hr.- 1 d' 4 Except for detector calibration, which refers onlyf to the^ use of a 100 mr check source, the test procedure appears-to be sufficiently detailed and adequate to demonstrate this subsystem. Mr. Kaulbach has stated his intention to develop a more detailed radiation calibration pro-i cedure in conjunction with GE representatives. This procedure will be reviewed further when the details on detector calibration are made available. t ': I "TFrg-D- y"v4db 'vPir ryevis Ty 7-yv v-- pme' - - - +m++c+- c e %> esc --~uv'=, amis -*a m A-= m-t
m--
d-M PP"
rc^ .a s W -; m A1h,_ _; _, _., ~ ~* l1f f%.. t ~~1 -t j % fe O - 56:- 1 .i '5. (D-17-A) Off-Gas' Air Ejector Monitoring Subsystem rand a (D-17-B) Stack Gas Monitoring Subsystem l I 2 l l o The stated purposes of these. procedures are discussed in J. CO Report No. 219/67-6,-paragraph III.C.. j "m 1 , [' Both monitoring subsystems. employ two channels'of'in-l ^ strumentation. The Off-Gas system has-two ion chambers, j each "looking at" the same special section of pipe:in.the? A off-gas sample line upstream of'the-system absolute. filters. The Stack Gas' system has two scintillation detectors j monitoring the gas sample after it has passed.through a particulate filter and a -halogen filter. The filters l are removed periodically and monitored for particulates
- j and halogens..
j l Mr.- Doyle has indicated he will supplement ~the preoperational~. j procedures by developing more comprehensiveL etector cali-l d bration procedures. These will be developed with GE assistance. Additionally, Mr. Doyle will develop. pro-1
- p cedures for in place testing of filters employed in-these 5
' monitoring. systems. These test procedures will be reviewed-further when this additional' detail becomes available. j 6. (D-18) Environs Monitorino (Fixed Stations on S'ite) a The stated purpose is to. demonstrate that all environs monitoring stations (fixed stations on site) are; operable, in calibration, have correct' trip settings and perform i . [ specified annunciation functions on high or low trip. i The three effluent gas monitors used at each of=the three-j fixed stations have a dynamic range of'O.01 .100-mr/hr. I i Detailed procedures for detector calibration have not been4 1 i[ incorporated in these preoperational procedures. Except l for this, the test procedures appear' to be sufficiently-detailed and adequate to demonstrate the subject equipment.. } Mr. Kaulback intends to develop' detector calibration -pro-l cedures' and procedures to calibrate and determine sample l collection efficiency for the particulate sampler. Pa r-ticulate samples will be removed for counting and evaluation. i 1
p~ , p., o 3 ,r 5 4 VI. Exit Interviews Exit interviews were held with pertinent JC and GE repre- .g sentatives at the conclusions of some of the visits discussed j in this report. In other instances, exit interviews were not f required due to the nature of the visit or because the persons I involved were in the company of the ' inspector (s) for most or all '. I of the visit. In all cases, the pertinent' items discussed and f the significant comments made by those interviewed are contained 'j within the body of the report.. t k ( t I I +,% 1
f_/ c.r r., - ~ ~,. - ..A;*
- ,4. n ' :e'.
i .s (i; r,.=. ). - i,. .: f. s,. v-4 JERSEY CDTTRAL POWER & LIGHT COMPANY (CO REPORT NO. 219/68-1) r.. -I .i ADDISTDUM I ,'x w ;l, f %r 1 i k. l i ' l 1 l O l f 4 f' y. t i a I l l 1 l I I I i. O u =, eww m__.h m_.-
,. ~_.-... g;. m
- a
.cr. .,g,=.-- .. :..u. .w. . s. .._. u,. - ,(;o eu - * - l
- f. _, j p qy..
o 9,.. .;p );' i .L - t c I: 1 e t.. 2 JERSEY CENTRAL POWER & LIGHT COMPANY u (CO' REPORT NO.- 219/68-1)- l. .. 4,,- y,'. ( P L. l'.3 ADDENDUM II .a :m l: QQ;
- N
(( ^t.y. 5 i 4 6 h 6 6 f f i p i i a [ .1 'j ? S 3 4 9 6 m l h 3 ^ ' " " "
- r3 w-
'r g- ,-.a.rrAw .~. g 9+ 7 .y h-e--.m ,9-. ' g
p, 7 a (( ly.;".Q ~ ~ - , ;.L. f h .}
- t.f,
- .
7-i
- c, j:
e. a . p..f. ' JERSEY CENTRAL POWER- & LIGHT COMPANY (CO REnfRT NO. 219/68-1) J: - ig ' ADDENDUM-III w .' g INDEX - STARTUP TEST PROCEDURES 3 j- , a i .. 6
- - i 1.
Chemical & Radiochemical Tests andL Measurement!s ) s t 9' 2. Control-Rod Drive System 3. Fuel Loading f 4. Shutdown Margin 5. . Radiation Measurements 6. Vibration Measurements 7. Control Rod Sequence 8. Startup Range Instrumentation - SRM i 9. IRM - Intermediate Range Instrumentation o 10. Reactor Vessel, Temperatures 11. Sycter. Expansion .7 + 12. Main Steam Isolation Valves i t j'- 113. Isolation condenser System 14. Recirculation System Performance q 15. Flow Control-I 16. Primary System Relief Valves ] 17. Turbine Trip k 4 ,m
f; '~ '; :,..a.__4 a_ pv, c,,e 4..,, v. e. sj. lh,; } ' J,: 1 -i 18. Generator Trip. g 19. Pressure Regulator ,4 20. Bypass Valves 21. Feedwater Pumps
- +.Jd-22.
Flux Response to Rods 1 [ 23. Local Power Range Monitor Calibration - LPRM 24. Average' Power Range Monitor Calibration - APRM 25. Core Performance Evaluation 26. Power Calibration of Rods 27. Axial Power Distribution Measurement 28. Rod Pattern Exchange 29. Steam Separator - Dryer Performance 30. Electrical Output and Heat Rate A. Transient Recorder Program B.. Recorder Signal Taps Index h C. Standard Procedures } _... ~.... -+ .}}