ML20087A296

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Co Rept 50-219/69-01 on 681205-690206.Rept Summarizes Status of Resolution of All Outstanding Issues
ML20087A296
Person / Time
Site: Oyster Creek
Issue date: 02/17/1969
From: Bryan S, Robert Carlson, Gilbert R, James Keppler, Nolan F
US ATOMIC ENERGY COMMISSION (AEC)
To:
Shared Package
ML20086U000 List: ... further results
References
FOIA-95-36 50-219-69-01, 50-219-69-1, NUDOCS 9508040263
Download: ML20087A296 (52)
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U. S. ATOMIC ENERGY COMMISSION REGION I DIVISION OF COMPLIANCE I

Report of Inspection CO Reporti No. 219/69-1

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Licensee:

JERSEY CENTRAL POWER-& LIGHT COMPANY

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(Oyster Creek 1)

- Construction Permit No. CPPR-15 l

Category B Dates of Inspections:

December 5, 6, 17 and 18, 1968 January 14 - 18, 23, 24 and 29 - 31,1969 February 3, 4 and 6, 1969 Dates of Previous Inspection:

December 10 - 12, 1968 b"

Inspected by:

1 R. T. Carlson, Senior Reactor Inspector Date 1m Ca~ A 2W F. J. Nolan, Sr. React'or Inspection Specialist Date I'lfW}

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S. B a, Reactor Engineer

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'R. G. Gilbert, Radiation Specialist Date

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Reviewed by : 4Ab 1/n n

8. G. Ke%#1er, Senior Reactor Inspection Date Specialist Proprietary Information:

None I

SUMMARY

The Division of Compliance's construction phase inspection program for the subject facility is essentially complete.

This report summarizes the status of resolution of all outstanding issues as of the last visit in the subject inspection' period.

Those issues considered by Compliance to require resolution before a recommenda-tion can be made to DRL relating to the issuance of the provisional operational license are summarized below.

References to the pertinent sections in the body of the report are given in parenthesis.

9500040263 950227 PDR FOIA DEKOK95-36 PDR

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Adequacy.of investigation and resolution of valve crack problems. -(II.A.4) f

2.. Adequacy of evaluation of cable tray loading and plans ' for in-plant. verification.

(II.A.3)'

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Adequacy.of. calibration of effluent monitors - demonstration j

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of proper isokinetic sampling.

(V.H.2)

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Adequacy of GE's re-evaluation of containment penetration.

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Item referred to DRL for further consideration.

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Adequacy of nondestructive testing of' safety valves..

(II.A.7) i 6.

Satisfactory 1 completion and evaluation of outstanding pre-1 operational testing.

(III) 7.

Adequacy of operational environmental monitoring program.

Item 1 I

referred to DRL for further consideration..

(II.D) 8.

Conflict'in Application relating to check valve in containment

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spray system.

Item referred to DRL for further' consideration.

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(II.G) l

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Completion of outstanding construction items in accordance with agreed-to proposal by GE.

(II.H) j 10.

Adequacy of final approved core-loading' procedure.-

(II.B.4)

11.. Adequacy of final approved startup testing program.

(II. B. 5 ) -

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- 12. - Satisfactory completion of preparation of plant operating A

procedures and resolution of identified deficiencies.

.(IV.F) l 13.

Satisfactory completion of CO review of JC-GE operating.

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organizations for determination of conformance with Application-

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and readiness to assume operating responsibilities.

(II.I) l l

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Satisfactory' demonstration of emergency evacuation alarm.

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(II.B.7.d)

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Question regarding open sleeve in watertight compartment bulkhead.

(II.C.14)

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Scope of Visits 6

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The Compliance construction phase inspection program at the Jersey.

4 Central Power and Light Company's Oyster Creek l facility was continued.

Inspection visits were performed during the period covered by this report as follows:

Carlson (Section II)

January 14-16 and 29-31, 1969 Nolan (Section III) -

December 5, 6, 17 and 18, 1968 January 17, 18, 23, 24, 30 and 31, 1969 Bryan (Section IV)

December 17 and 18, 1968 January 30 and 31, 1969 Gilbert (Section V)

February 3, 4 and 6, 1969 Mr. D. L. Caphton, Reactor Inspector, CO:I, accompanied Mr. Carlson during the Janoary 14-16 and 29-31, 1969 visits and contributed to Section II of the report where indicated.

The principal persons contacted during these visits were as follows:

Jersey Central Power & Light Company (JC)

T. J. McCluskey, Plant Superintendent D. E. Hetrick, Operations Supervisor I.

R. Finfrock, Jr., Technical Supervisor D.

Ross, Technical Supervisor Trainee N. M. Nelson, Maintenance Supervisor R. D. Doyle, Chemical Supervisor

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D. E. Kaulback, Radiation Protection Supervisor k

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E. J. Riggle, Electrical and Instrument Foreman General Electric Company (GE)

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D. K. Willett, Project Manager (San Jose)

A. G. Mellor, Electrical Engineer, Balance-of-Plant Engineering (San Jose)

D. E. Tackett, Field Application Engineer (San Jose)

N. M. Strand, Construction Site Manager R. C. Green, Test Engineer R. M. Haynes, Lead Construction Engineer D. L. Borchers, Electrical Construction Engineer R. Blundell, Mechanical Construction Engineer i

W. C. Bibb, Operations Superintendent i

L. Sandlin, Scheduler I

Burns and Roe (B&R)

G. A. Lari, Project Engineer (Oradell)

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  1. II.'LResults of Visits'- Carlson j

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Plant Quality

~ Status of Outstanding Issues-t O

.The. status-of. resolution of-all outstanding plant' quality.

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deficiencies as of the last visit covered by this report:is.

summarized below.

For purposes of ~ comparison, - the pertinent

. i references to previous inspection reports, where applicable, are given in parenthesis.

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Electrical Cable Separation'(219/68-10, Paragraph II.A.18, E

4' Paragraph II.B.2 and Paragraph II.D.)

The inspector determined that the correction of the deficiencies identified in this subject area by-both GE l

and CO in their respective audits had been satisfactorily l 1

completed.

This determination was accomplished through a review of field engineering records, discussionsLwith

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cognizant field engineers, and a sampling. audit of the

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physical installations by the inspector in the company of i

Mr. Borchers.

For ' record purposes the following' statistics, -

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obtained from the site records, reflect the magnitude of this action:

Total No.-cables pulled 44 j

Total footage of new cable 17,125' Total footage of-new conduit 1,010' Total footage cxE new tray -

16' Total No. concrete borings (For new access routes) 2 l

l l-The inspector also determined that GE had properly.

coordinated this corrective action with the preoperational testing such that where appropriate the'new installation would be included in the test program.

I l-Based on the above, this matter is considere'd to be -'

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resolved.

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Battery R'oom ventila' tion- (219/68-10, Paragraph II.A.21):

' The 'following additional -informationf regarding the ability-

. of the subject system to cope with hydrogen in a ' situation ~

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. ventilation. air. flow loss was obtained-through_ discussions with Messrs. Loeb, McCluskey, Hetrick and Lari-(by phone):

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a.

.The design of the 125' volt D.C.' system'is such that-the'M-G se,ts and battery chargers; serve the normal-1 system needs.

The batteries fulfill this. function when all other sources fail, b.

To get into a situation of hydrogen explosion potential

would require:

A-loss of pwer that" would deactivate - t both' fans and the battery chargers;,an extensive period (>7 hours) - of battery discharging without:

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restoring.even emergency power to the battery chargers;;

and a like-period of ; battery charging without 're-storing. power to the ventilation system _' fans (a total-of 714 hours0.00826 days <br />0.198 hours <br />0.00118 weeks <br />2.71677e-4 months <br /> without pwer).

c.

According to Mr. Loeb, in zthe worst case situation (total loss of. power.with ' loss of: coolant ' accident maximum demand on batteries), the period of peak load demand on the batteries would be over in one -to two hours. -In this situation, the greatest load demand comes from the turbine-generator emergency lube oil-

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pumps, which, according to Mr. McCluskeyi would j'

normally be shut down within 1/2 to one hour following.

the loss of power incident... When the' inspector pointed out that this period would be a trying one on the.

operators and that shutdown of the lube oil pumps might easily be. overlooked, Mr. McCluskey stated that:

a requirement for completion of this action would.be

'I incorporated in the appropriate emergency procedure..

d With-regard to the restoration of power to.the battery-chargers, the inspector determined.the facility r

emergency procedures for. loss of-power -include a

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f requirement for restarting this equipment immediately

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j emergency power (d'iesels).

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The emergency procedures do not specify the restora-4

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tion of power to the battery room fans while on. the 4

emergency power source: however, the procedures do call for restarting the fans following restoration of' L

normal power.

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In discussing the situation postulated above,. Messrs.

McCluskey and Hetrick both stated that the probability l

of experiencing the combination of events, including procedural violations, necessary in order to get into-l a situation of hydrogen explosion potential (even dis-l regarding the benefits of natural draft in the battery 3 room) was too low to be of concern.

Based on the information presented above and in the referenced report, it appears that the subject ventilation system, although not as conservative in design as might be desired, is adequate for this application.

This matter is considered resolved.

3.

Cable Tray Loading (219/68-10, Paragraph II.B.3)

GE still had not made the results of their investigative work available for CO review as of the visits covered by this report; however, they stated that it would be avail-able for review during a CO-JC-GE meeting scheduled to be held at the site on February 12, 1969.

This matter remains unresolved at this time.

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4.

Valve Crack Problems (219/68-10, Paragrabh II.E.)

j The status of GE's investigation and corrective action as of the visits covered by this report is as follows:

i a.

A total of 43 valves had been examined.

All of the f

valves came from the same vendor, Anchor Equipment Co.,

I representatives of which have been involved in some of f

the field investigation work.

Systems involved

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/:. J included the. core spray (8), shutdown cooling (2),

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isolation condenser (8), cleanup demineralizer (24),

-O' and control rod drive return line (1).

The type inspection performed and number of valves' involved fol19Ws:

t Type No.-Valves 3

t outside visual 36 Inside visual (10 X Glass)

'8 Outside PT 28 Radiography (In Part or Total) 5 In addition to the above, the vendor radiographs of the original group of 21 valves examined plus-five others where this type followup was deemed advisable are being reviewed by GE quality control (Mr. Tackett).

b.

Replacement of core spray valve V-20-41 is underway.

The metallurgical examination'of the original valve is incomplete.

c.

A total of 16 valves have required at least some grinding.

d.

Four valves require significant repair.

The valves,

maximum depth of grinding, and current status are 3

given below:

l Max. Depth 1

Valve System Grinding, Inches Status

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V-17-19 Shutdown Cooling 3/4 Repairs 5

(Inside Drywell)

Complete V-17-54 Shutdown Cooling 1

Repairs (Inside Drywell)

Complete V-14-34 Isolation Condenser Return 1+

Repairs (Outside Drywell)

Incomplete Repairs V-20-17 Core Spray (Inside Drywell)

Incomplete

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Regarding valve V-20-17,.the review of the vendor; i

Ya-radiographs revealed the presence;of a 3-4" shrink

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crack in'the interior structure-of the valve casting.

31 This condition'which was cor*irmed'by reshooting in-l the field'was.not visible otherwise.

GE is continuing-J' q'

its investigation of this. valve..-

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may require replacement.

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The special double isolation valves in the isolation condenser system inlet lines were. included'in the 28 valves. d y e penetrant tested.

A total of 18 areas--

i were found to require grinding,- several to a depth 'of l

1/4".

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Mr. Tackett' stated-that he was attempting to correlate the indications detected in. the. field with those noted-on the' vendor radiographs. _He said that;those detected i

on core spray valve V-20-41 :(replaced): were shown on the radiographs.

In response to a specific' question,.

Mr. Tackett stated that one objective of'his-review of-the radiographs was to determine if there had;been'any, propagation of.the indications observed.

He stated'that no evidence to that effect had.been. noted thus far.

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In response to a question regarding the cause of the conditions noted, Mr. hckett stated;that it was.his opinion'that the overall problem was. attributable to l

J the vendor limiting the radiographic review to one r

man.

Also, that it. appeared that the dye penetrant' 3

techniques employed by the vendor left-something to be' desired.

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Linear indications have F.lso been detected-on the re--

i circulation system pump suction valves (Chapman) in J

loops B and C.

In one instance (B), the indication.

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was found during the cleanup removal.of Nelson studs

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  • CO Report No. 219/68-5, Paragraph II.A.2.

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In the second instance - (C), the indication was found -

during cleanup of an' arc strike.

According. to Mr.

. Strand, the. indications,'which became. visible.only-l

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after they had ground.past the chilled skin of:the.

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,i-pump material, are of the type. internal discontinuity of-which a certain amount are normally. experienced-in; ~ he casting process.. He-said'that as such they were t

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acceptable to the-code;until they became surface. 3 indications as a result of their grindingLto remove: ' .i i the previously identified-defects. Visual-examination. 1 of the work by Mr. Caphton on January 31, 1969, showed - i the linear indication on the'C-loop valve to.be: 1-1/2 2" long. During. a phone communication to the site on February 1, 1969, the inspector learned that removal of the ' subject - l indications and application of' repair weld.had been-. t completed on January 31, 1969.. According-to Mr. Strand, this was accomplished within the maximum amount .l permissible beyond which radiography and rehydro would 'j be required. He stated that the engineering groups l from GE and the vendor participated in this determination. A meeting between CO-JC-GE representatives has been .j scheduled at the site :on February 12, 1969, to discuss j further the cause and significance of the valve crack-problems, including those in the recirculation. system. j Mr. J. Collins, Metallurgist,-COiHQ,.Dr. Gilliland,- P_arameters, Inc. (CO consultant) and Mr. R. Gustafson, DRS, q will participate in these discussions. This' matter remains [ outstanding pending adequate resolution.of these problems. .i 4 5. Core' Spray Nozzles - Question on Adequacy of Type Material f Used ~ i -p The subject nozzles are made of type 303 stainless steel. j Similar material is in use at Nine Mile Point. During)a .I previous visit, the inspector requested.Mr. Strand to i determine. the suitability of this material in. this 1 'l I + - ~. -.

~ ? ' [if yy ^ X:,., 0 3 y r -._ 11 _ a W . f,. 4 d -application in light of the.known difficulties in welding-j L. (requires special heat treatment). Mr. Strand reported.~that. ~ il GE engineering re-evaluated the situation and they.saw no d -problem. He stated that'the nozzles are screwed in and are l N tack welded only. This matter.is' considered resolved. L g 6. -Recirculatf.on System Valves and-Lines Applicability i l of Problers Detected at-Nine Mile Point (NMP)- 4 3 1 Discussicns.with.Mr. Strand regarding.the subject' problems L laps and. slivers.in the piping, slag in valve gate contact surfaces and cracks in the stellited surfaces of 5 the valves - L ' revealed the following information relating;to this facility: l a. GE at this facility was aware offthe problems experienced at NMP. b. The subject equipment here. was inspected and-the necessary corrective actions. completed prior.to the hydro performed.following the pressure vessel repairs (stub tube cracks, etc.) c. The inspection and repair work was performed by the ~ same GE and vendor representatives involved at NMP. d. With the exception of the cracks in the valve stellited surfaces (none noted here),, conditions similar to the others detected at NMP wore noted at this facility. b e. Per Mr. Strand, all deliterious effects.were removed from the valves. Regarding the' pipe' surfaces, he- ) stated that some of the defects were removed.- The I remainder were determined to be adequate as is by GE engineering and quality control personnel. 4 Based on the above, this matter is cbnsidered to be resolved. 5 E- [ ( i ...u.,.mn-w..,--.nn+- ~. - - - - - ~ ~ - ~ ~ ~ ~ - ~ - - - ~ ~ ~ - ' ~v -~~~ ~ ~ ' + .----6 --i

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.During;a previous meeting between DRL-CO and'JC-GE*, j j~j held for the. purpose of discussing the valve crack problems ,j at this facility, it,was ' learned that the purchase y ' specifications for the: safety valves did not, include.a' ,1 y requirement for radiographic examination'of the valve m4 castings..The'following additional information was l obtained 'in subsequent-communications with Mr. Loeb: ' l a. 'The valves are composed of a cast carbon steel e body with a cast stainless' steel insert screwed into' ) the body. The carbon steel casting was magnetic. particle tested following final heat treatment. The i stainless steel portion was dye penetrant tes'ted_~ l following heat treatment. Per Mr. Loeb,..the--latter E piece was smaller than the minimum size above which I the code requires radiography; therefore,-none-was performed. 1 b. 'Per Mr. Loeb, the carbon steel' casting does not'see. system pressure.** In addition, the' valve is of a 1200 psi or above pressure rating and as.such hasLa: 1 large designed-in safety margin. He stated that on-J the basis of these facts, the engineers. decided that ~ they could waiver any radiography of the casting. j Additional information is to be obtained on this subject.

l during a visit'to the site. scheduled for February.12:- 14,.

j 1969. This matter remains outstanding-'at this. time. .l r i i

  • Held at Headquarters on January 17, 1969, and attended by the.

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    • Additional information provided by Mr. Loeb would seem to_

1 ] I indicate that this is not true of the flange by which'the valve is attached to the main steam header. 1 E ..,., _.. -..,... ~. - -.. -..,..,.. - -.. -.. - - -.,

97 y,. l 2.._--- ( j "3pp. u s' N- <+ ~ } , '; j _ i.' 4* l Q[ 7 n up- -B. Procedures -- Status of ' Outstanding Items a (( The status.of resolution'of all outstanding issues. relating-j + to preoperational and operational-procedures as of the last 1 S visit covered by this report is - summarized below. Rr purposes- -j i: of comparison, the pertinent references to previous inspection 4 .D _'1. reports', where applicable,.are given in parenthesis. j v

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-l '"j 1. Preoperational Testing - Auto Depressurization System i D i. (219/68-10, Paragraph II.G.2)' i t A test program has-been decided upon by GE. The. testing will include a: demonstration of the proper operation of the control circuitry (without operation of the valves), i followed by manual actuation of the electromatic relief valves, one at.a time. The latter tests,;to be performed-l during the cooldown following primary system-expansion: measurements, will result in actual limited blowdowns to the torus. Approved test procedures were determined t'o > have been generated for-this work. The above' described test program-for the auto depressuriza-tion system is considered acceptable to CO. This matter -is considered resolved. 2. Preoperational Testing'- Radiation Monitoring Systems (Area, Process, Off-Gas and Stack) (219/68-2, Paragraph II.G. and 68-1, Paragraph V.F.) .l The results of co's review' of the : subject procedures are { discussed in paragraph V of this report. f i 3. Integrated Functional Test (219/68-8, Paragraph II.D.l.)- I 'E The approved, detailed procedures for this test were =! ..i-reviewed by the: inspector and determined to be adequate. This matter is considered to be resolved, t 4. core Loading Procedure (219/68-8, Paragraph II.D.2.)- l q The final, approved procedure was still not available for CO review at the time of these visits; however, 1 GB-JC indicated that it would be available for review l during the visit scheduled for the week of February 10, ) 1969. This matter remains outstanding at this time. i ) 7

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Startup Testing Program ( 219/68-8, Paragraph II.E.): b4 y< i, Tha final, approved procedures were still not'available i "' for CO review at. the -time.of these visits. This matter-remains outstanding at.this time.

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, ] 6. Plant Operatino' Procedures (219/68-2, Paragraph' II. I. ) u ~ 7.( The-results offCO's review of these proceduresfis-y; discussed in paragraphs IV and v.of this report.. 1 ^ 7. Emergency Plans (219/68-1, Paragraph II.C. and'68-2,- Paragraph II.4. ) * \\ ~ The emergency plans for OC-1 were rev1ewed Jin comparison: I with the commitments in the Application ** and with the U criteria specified in Co Manual, Chapter 2005. i The results of the review showed that.pending theilicensee'si final approval of the Emergency. Handbook- (scheduled to be ] completed prior to fuel loading), the plans are in j conformance with the Application. =l t Personnel orientation in'the plans has begun and is .j scheduled to be completed prior to the initiation of power operation. This is considered adequate by the: inspectors. No date has been established for the initial' evacuation. drill. This matter will be followed-by the' 1 inspector as is appropriate. 7 l -t i Y'

  • This section of the report was contributed to by D. L.!

i ) Caphton. J

    • Amendment No. 11, Page VIII-5-1 (with attachment); Amendment No.

) 21; ' and Amendment No. 41. y +-

vy..,. 3&g,.lj') k"j ^ S~ "^~~'A-~"" + ~ l ., N M * / 3. ' {.k ;-f }, l .2-t - 15 l, 1 - ;p, .i The following. discussion relates-to-the areas in the' <j~ OC-1 emergency plans that' were divergent with -respect. 3 a .to the-CO objectives and requirements : set down in the '. ~] - Compliance Manual. - The specific areas 'of: contention and" l the JC comments relating.to each areai(provided by' ij,' Mr.. Finfrock). are given. in the following 'sutunary. 'Each. Y item is referenced to the appropriate paragraph of,the. 1 g. Compliance Manual, Chapter'2005. i .h a. Establishment 1of Authority and Responsibility. ) (Paragraph 2005-03'a.l.) The procedures ~were not clear as to decision-making j authority during.an emergency-in the event a conflict I of interest occurred.between the control. room foreman: and the emergency control station (ECS). supervisor. - 4 Ans. - The control room foreman has - the burden' of j responsibility and authority for the' entire plant'. while - on duty,-- therefore he would have the ultimate authority and decision-making responsibility. Li ^ b. Availability of Information for continuing Evaluation of Situation - (Paragraph 2005-03 d.2.) j j The inspector asked about the~ capabilities of the.ECS' in a situation where the OC-1 control room was to become uninhabitable, i Ans. - OC-1 has no secondary control room.- Information -j i. is only available to permit; continuing' evaluation of' emergency situations from the' control room. It is 4 JC's intent to man the control room.throughout-all l .j emergencies. The emergency plan,therefore is based j on the continued habitability of the control room. 4 l' (Paragraphs 2005-03 e.3 and e.4.) - 4 c. Action Levels i The plan does not provide action levels.in detail I commensurate with an emergency progressing from.one degree of hazard to another. j q t -...m.-g. ,,... ~,,,.,. e,.n..-.n n - - ..a ,.w.

N tv @iM ' "' ' - 1- - - - -- - ~-:- - ^ Pg jf p q RJ J m;. .u, i. S Vi L< j, ' -x y Ans. - Decisions on responsive action to emergencies i a is left_to supervision. d. Evacuation'or Take Cover Signals (Paragraph 2005-03 c.2.)! 5f ), si The inspector asked if the station' evacuation alarm had nf been tested and if personnel were familiar with the.. Mr signals. p Ans. - The alarm had not been tested to date, nor was-it known at this time as to the type of alarm provided. CO will followup on this matter. e. Evacuation Routes (Paragraph ' 2005-03 q.3.) No primary evacuation routes' had been specified at-the time of the visit. The plans for primary-and secondary routes were questioned. Ans. - Only primary routes will' be specified. The signs will be installed at the conclusion of construc-tion activities. Secondary routes are not being specified in order to keep _the plan simple, f. Personnel' Accountability (Paragraph 2005-03 c.4.) Accountability for personnel following'an emergency is not covered in detail by procedures. Ans. - It is the responsibility of line supervision ~ g to provide accountability for all of their people. 4 Visitors will be under the direct supervision of trained personnel who will provide for. their accountability. l All personnel on the OC-1 site therefore will be in the accountability _ system of line supervision. ~ g. Site Access (Paragraph 2005-03 c.5.) ~ JC was.' asked if alternate access gates were available in the event a radiation cloud were to descend upon the normal gate.

g .y - g,.. f ;.,b ~, .; g :: ' p q M.; () () i f 7.[ .- 17'- .fi ~ q Ans. - No1 alternate access gates were specified. [ r The calculated dose rates based upon'the design 'l } accident are relatively low. h. Emergency Equipment and' Reference Material' (Paragraph 2005-03.h.l.) m p Drawings concerning the plant were not listed in g g' the ECS equipment list. t-; Ans._- Drawings are available from the Parsippany. -l office if needed. A complete. set of plant specifica-tions, etc., are available from that office. ) ~ The-inspector asked about the availability of keys to provide access to the emergency control' center. Ans. - Keys will be made available to supervisory personnel. JC has company rules covering. entrance of f personnel to switch yards where the ECS is located. Special rulings may be required by JC to permit. access I to the ECS. CO will provide followup on the key distribution question. 1. Guidelines for Maximum Emergency Exposures (Paragraph 2005-03 1.2.) The plans do not specify specific guidelines for - maximum emergency exposure of personnel.. i i l Ans. - The guidelines were intentionally omitted in -l "j order to avoid any possible legal = conflict with 10 CFR l 20 which is ' specific in allowable permissible expo-i sures to licensee personnel. j j. Guidelines for Emergency Re-entry ' (Paracraph 2005-03Li.8'.) Specific guidelines governing. emergency re-entry are f not covered in the plans. 1 i Ans. - It is felt that there,would be adequate decision-making time after an emergency based on the design accident to determine re-entry plans to suit. -l l i s - ~. - ~. i

nb hh h ~ ~" " ~ "" " -~ ~~~~~ ~ ~.m n Y ll p) ,m 0: --{f - -P ,y .v .k. Practice Drills (Paragraph 2005-03~1.4.)

4..

( The. plans did not specify a frequency of practice m drills. 1 [ Ans. - Practice drills will be held such that all .l station personnel will participate in one drill '#f annually. ] C. Status of'Significant Items Identified in DRL Report:No. 4 to ACRS* The information summarized belaw was 'obtained through a review of field drawings, a visual examination of' installed equipment and discussions with cognizant JC-GE personnel. For comparison purposes, the pertinent reference in t he subject ' report. is - shown in parenthesis. 1. A second monitor is installed in the liquid effluent discharge line. (Paragraph 3.0, Page 7) l_,, 2. The post-operational environs monitoring program planned by the licensee is not in complete accordance with that implied in DRL Report No. 4. This subject is discussed in more detail in paragraph II.D. of this report. (Paragraph 3.0, Page 7) 3. The control rod velocity limiters have been installed. (Paragraph 4.2, Page 12) i l 4. The automatic recirculation flow controllers are installed-j but are not operational and are not scheduled to be tested at this time. (Paragraph 4.3.3, Page 15)' 5. The design injection rate for the standby liquid contro1L 4 i system is scheduled to be verified-during preoperational~ testing. (Paragraph 4.3.4, Page 16), h 6. A check valve has been installed downstream in the liquid poison pump discharge circuit of each relief line. (Paragraph 4.3.4, Page 17) f'.

  • Jersey Central Power & Light Co Oyster Creek 1, Report No. 4 to

'ACRS, Docket No. 50-219, dated October 14, 1968. .v%ugu. -*s---gi -w-* 4-mm a w=-4 a 'o%-~ _. g

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- 7.- A separator type accumulator-is. installed on the outlet 1 of each liquid poison. pump. (Paragraph 4.3.4', Page>17) ~~ i

8..

A. total of 16* safety' valves'and four relief valves are-q provided.- (Paragraph-5.2, Page 20) ? ' In the' primary system leak detection system, the. p~s@ 9. condensate from the fan coolers -is routed to the floor - j drain sump,.and alarms'are installed ~to. indicate that the flow into the sump or drain tanks is exceeding a, 2 present rate.. (Paragraph 5.3,~Page 21) 10. In'the. isolation condenser system, concentric guards pipes. are provided in each steam supply'line to extend the contain-- L~ ment boundary-to the first valve body of the special double isolation-valves. (Paragraph. 5.6.1, Page 24): ~ 11. The startup test-program specifies that the primary contain-ment atmosphere will be inerted with nitrogen upon completion-I of the initial power test program. (Paragraph 6.1.1, Page 34) 12. The preoperational leak test procedure for the primary. containment has been modified such that the new initial test pressure for both the drywell and absorption chamber-will be 35 psig. (Paragraph 6.1.2,.-Page 35) 13. The standby gas treatment system is being modified to provide continuous air flow through the filters. l 1 (Paragraph 6.2.1, Page 44) .l [ 14. The following modifications have been implemented to accommodate a passive element failure in the BCS. 1 (Paragraph 7.4.4, Page'62) Y a. The fire water system has been connected'to the l-core spray systems.

  • DRL Report No. 4 references 15 safety valves; however, this'

,L number was increased to 16 in Amendment No. 45,-issued subsequently. 1 y~,-.,.-.- n -- . ~. I~

em- - ____...._.-...a_. -____a -. s_.. - 3 )l,. ) k) / d-j; O

b. - ' Penetrations to the pump compartments have been l
sealed, c.

Water-tight doors have been installed at the entrances. L ~ " to the' pump _ compartments. l: l .d. The. drain lines from the sumps have been provided with valves to prevent backflow. "El Regarding item c, an open sleeve was noted to be installed I through one of the door frames, at an elevation of 86" above the torus room floor. This is.less than the anticipated 8 ' - (96") flood level specified in the referenced section of-DRL Report No. 4. GE has been requested to justify this installation. This matter remains outstanding at this time. 15. Flow restrictions are installed in the main steam lines. These are physically located in the vertical downcomers between the safety valves and the inside drywell isolation l valves. (Paragraph 8.1, Page 75) l 16. The control rod housing support structure was-installed but has since been removed temporarily for accessibility during control rod drive testing. CO will confirm its re-installation at the proper time. (Paragraph 8.3, Page 76) 17. The elevated water storage tank has been removed. (Paragraph 10.8, Page 96) l 18. An automatic isolation valve has been installed on the discharge side of the condenser vacuum pump. The valve will be closed by a high radiation signal from the steam line monitors. (Paragraph 12.1, Page 106). I D. Post-Operational Environs Monitoring Program l Jersey Central's planned post-operational environs monitoring -program was reviewed with Mr. Finfrock.. The program differs significantly (mostly deletions) from the preoperational program and from that implied in DRL Report No. 4 to the ACRS.* A copy of the preoperational program ** marked up to reflect the~ post-operational program, is included as Attachment A to this report.

  • See Paragraph II.C.2 of this report.
    • FDSAR, Volume I, Table II-6-1, Pages II-6-2 and 3.

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9 1: 21 -- i [t I 4 d. i The above observations were subsequently discussed in detail ,q with Mr. Tedesco, Chief, Reactor. Proj ect Branch 2, - who1 following;an evaluatien of the changes noted, informed the j ,'i writer that.the program as described was not in keeping with j the understandings that DRL had.as a result of their. l Q communications with the licensee. in this '. regard. The' writer' j "' subsequently informed Mr. McCluskey.of thisl fact. 'Mr. McCluskey's yfp. response was that the planned program was c'onsistent with their' ~. ' commitments. This matter. remains _ outstanding at this ' time. i 4-E. Fire Protection System

  • The inspector completed a site inspection of the' fire protection system during the January 14 - 16 and 29 - 31 visits.

The inspector visually checked the component parts of the.' system-against the commitments stated in the FDSAR Volume I, paragraph X-3.1. The system was found to be installed per the FDSAR;With . j oneiexception; the FDSAR stated "An. overhead water' tank.' floats j on the system and provides potable water and also provides l 85,000 gallons in reserve for fire. fighting should the diesel ) pumps fail." The overhead water tank has been disconnected. ' from the fire protection system and-removed from the siterin-accordance' with commitments made in Amendmeri^t No. 28, paragraph l III.A.3. The inspector observed that some equipment has been installed that is over and above the requirement of the FDSARr for example, heat sensors in the hydrogen storage. area and turbine oil reservoir area, and alarms on the' control room. panel-to.- r signify a. fire pump is operating or malfunctioning. In j addition, Grinnell sprinkler fire. alarm bells are located'at the main fire protection system' supply. headers. Portable fire j extinguishers'are located throughout the building. ' i The fire protection system had not, at the' time of the j 1 inspection, been turned over.to Jersey Central by GE. The -water sprinkler system had main block valves closed. Hose stations did not have fire hoses installed..GE.and JC assured j the inspector' that this work would be completed prior to l initial core loading.

  • This section of the report was contributed to D. L. Caphton.

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93 + c - 22.- F. Containment Penetration Restraints _,m. Preliminary information relating to GE's.re-evaluation in-A ;. this area is discussed in CO' Report No.;219/68-10, Paragraph II.F.* Little new information was made available.to the. 3 inapector on-this subject until Mr. Loeb called the Region I 'I office on February 5, 1969. The pertinent facts,from'this Y latter communication are reflected below: p [i 1. The GE re-evaluation focused in on Design Criterion:I.B;* -- ~ l that relating.to the pipe rupture design condition.: The-other two criteria, those relating to the accident design condition and the maximum earthquake design condition, were not affected. 2. According to Mr. Loeb, the re-evaluation resulted in changes to the following statements in the~ referenced amendment: "For the loading on a penetration assembly resulting-e from a rupture of an adjacent pipe, the jet loading on the penetration assembly shall be defined as follows:- 1 x nominal line pressure of ruptured line x project l area on penetration assembly component-The above conditions of jet. force.. loadings do not1 occur [ simultaneously. For these conditions, the resultant' stresses in the drywell penetration shall not exceed- { 90% of the material yield stress of 33,000 psi." i p 3. According to Mr. Loeb, the approach given in item 2 I assumed a solid stream out of the open pipe, with no j allowance. f o r dispersion. He stated that the'new 4 approach takes credit for stream dispersion plus some ' [; flashing to steam. This results in the use'of~a l different formula from that.shown.above.; Mr. Loeb"was not'. l in a position.to communicate the revised-formula'to.the. inspector at the time. _Mr..Loeb stated that'this necessitated a change in the wording of the last. sentence in the above quote to something: approximating the following:

  • Amendment No. 11, Page III-15-1.

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?, ') ,I t 9 y L" - - - - The resultant. stresses in the drywell penetration shall : not exceed. 3 ' x Sm - (where - Sm = allowable stress for normal design conditions)." l l' 4. According'to Mr.'Loeb, implementation of the new approach ~ ^ at Oyster. creek 1 (which amounts to deletion of many restraints originally scheduled to'be installed inside the-F drywell) is essentially complete.. L 5. Mr. Loeb stated that the approach being applied to Oyster lL

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creeit 1 will also -be applied to all' other plants. 6. Mr. Loeb acknowledged that these changes do constitute a change in the Application; however, it was still within' the applicable codes and, therefore, did not necessarily require a. change in the public record. a The above information has been communicated to DRL for their a evaluation. This matter remains unresolved at*this time. G. Containment Spray System - Question Regarding Check Isolation valves The application contains-conflicting information relating to.- the presence of isolation check ~ valves in the containment spray system headers just inside the drywell as:is shown below: j 1. - FDSAR, Volume I, Paragraph V-1.6.1, contains the following statement: i' I " Automatic isolation is not used on the inlet lines ~ ] of the core spray, containment spray, and.feedwater - systems,.since operation of these systems is essential following a loss-of-coolant accident. Since the -l normal flow of water in these systems is Linward to the l. reactor vessel or primary containment, check valves-.. located in these lines, inside the drywell, will provide ~ automatic isolation.when'necessary." f 2. Tat V-1-2 in the above reference, which lists isolation: j valve requirements, does not specify check valves for the -j subj ect system. i .n ~~...-..--~~,.c- + 9 o., v. .,.._,y,,_,.

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'; f,,,. ~ - 24L-7 ~ 3. Amendment 15, Figure II-l ' (GE ' Drawing.886D325)', ~ showsl a "E check' valve. installed. - 1 g

  • 3 In discussing the above situation with Messrs. ' Loeb and Strand,
they each maintained that it wasinot intended that. check valves.

i be installed. A visual examination of the subject ~ equipment' fy. 'showed that no such valves were installed. This, subject has. @ g '- N'. been. referred to'DRL'for.,their. evaluation.* This' matter' l

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remains unresolved!as of'this time. i ,e . 'GE Proposal Relating to Status of Plantiat Time of-Core Loading j 5 H. At the ' request of - the writer, GE assembled a ? summary reflecting the proposed plant' status at the time of initilal' fuel loading.. l The study addressed itself to the completion. status ' of construc-1 tion,; construction testing and preoperational testing. The. l GE proposal has been reviewed -by CO and found to_ be acceptable. .l I. JC-GE Operatina Organizations 5 The inspector's review of the JC-GE organizations.to. determine conformance with the commitments in the Application and readiness to assume operational-responsibilities is underway - l and is scheduled to be completed during inspection visits scheduled for the weeks of February 10 and 17,1969. -This matter remains outstanding at this time. l t 6 l 1 I ] I 1-1 -i -IL f l

  • CO Report No. 219/68-10, Forwarding memorandum, Engelken to Boyd, dated February 7, 1969.

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l \\ 3 (J ,) III. Results of Visits - Nolan I ] This section of the report, which is concerned with the pre-operational testing program, was prepared by Mr. F. Nolan. i A. Scope of Compliance Review The results of the preoperational test program were reviewed with GE personnel on the dates indicated in paragraph C below. However, it should be noted that only four preoperational tests had attained the 100% completed status as of January 27, 1969. Following the 100% completion status, test results are reviewed by on-site GE personnel and then sent to GE-San Jose for further review by design personnel and/or reviewed via telephone conversation with appropriate design personnel depending on the complexity of the test results. In a number of tests, GE design personnel participate in and/or supervise the specific test. JC operating personnel make up 90% of the test crew; however, JC management personnel have not approved' and/or accepted any of the test results to date. Attached Table I provides information concerning the completion status of the preoperational test program on January 27, 1969. B. JC Review and Approval This writer reviewed the JC review and approval function for the preoperational testing program with Messrs. McCluskey and Hetrick at the conclusion of inspections on December. IB,1968 and January 1E, 24, and 30, 1969. At the December 18 meeting, Mr. McCluskey expressed JC's reluctance to indicate an acceptance of any systems or subsystens because of their contractual guarantee. The writer stated that JC, as the licensee, would have to at least informally indicate that all construction and preoperational testing, except AEC approved exemptions, were completed and that the facility was indeed ready for initial fuel loading. The writer also stated that this function 9?s normally accomplished by an in-house review and approval c: all test results by the licensee's proposed safety review or plant operations review committee. Mr. McCluskey stated that he would discuss this procedure with his management.

3 ) %) a During the January 18 inspection, Mr. McCluskey advised the writer that JC would provide the AEC with an informal indication of their review and approval of the preoperational test program and indicated that Mr. Hetrick would maintain a log of j informally approved test results. During the January 24 inspection, Mr. Hetrick stated that he was prepared to indicate approval of preoperational testing as discussed at previous s inspections. However, no test results had completed the GE and JC review programs at this time. During the January 30 inspection, Mr. McCluskey stated that JC could not perform the review and approval function described previously because of complications associated with the GE proposed plant completion status for initial fuel loading.* The problem was subsequently resolved during a telephone conversation on January 31, between Mr. D. Rees, JC, and Mr. J. P. O'Reilly, Chief, Reactor Inspection and Enforcement Branch. Based on this conversation, JC will perform an informal approval of the preoperational test program. ~ C. Preoperational Test Review Preoperational test results were reviewed during the inspection visits that were made on December 5, 6 and 17,1968, and January 17, 18, 23, 24 and 31, 1969. Specific tests reviewed during these inspections were the following: A-2, A-5, A-6, A-7, A-8, A-11, A-12, A-13, A-14, A-15, A-16, C-9, C-13, 3-2, D-7, D-8, D-13, and D-14. Prior to this review all pre-operational test procedures had been reviewed and approved by GE. However, this writer noted that JC had not completed a sign-off of these procedures as an indication of either review or approval. Details associated with the JC sign-off philosophy has been provided in CO Report No. 219/67-6, paragraphIII.B. t It should also be noted that none of the completed test results i had passed the GE and JC review and approval functions. The preoperational test results will be reviewed in detail during subsequent inspection visits.

  • Discussed in paragraph II.H. of this report.

l a ~

~ Y _._1 - T.._- i o 3 l' ? ; l D. Preoperational Tests Witnessed L core spray and control rod scram tests were witnessed during I the inspections conducted on January 23 and 24, 1969. e -J 1. Core Spray Testing core spray patterns within the reactor vessel were observed during one and two loop operation with various t combinations of main an.d booster pumps. During these i tests, data were obtained to permit the calibration of associated flow meters by monitoring known changes in reactor vessel level. The core spray patterns were monitored by both visual and photographic means. Normal operation will require one loop operation at approximately 4000 cpm. In addition to other patterns, this writer observed the spray patterns associated with 4000 cpm flow rates through each loop and a combined flow rate of 7400 cpm through both loops. Eacl. single loop core spray at 4000 cpm provided excellent coverage of 100% of the upper grid plate surface with a high density of water droplets. Based on these observations, this writer concluded that the single loop results compared favorably with photographs of tests that had been conducted at the GE-San Jose core spray test facility. This writer also compared photographs of the on-site test with visual results and concluded that the on-site visual results were superior to the on-site photographs because of back reflector light problems associated with this as-built facility. 2. Control Rod Testing As part of the review of the control rod drive system preoperational test, this writer witnessed a scram test i of a single control rod drive unit. During this test a multichannel events recorder was used to monitor the time required for scram operation following a simulated scram signal to the auxiliary power range monitors (APRM). The scram time was noted to require 1.5 seconds for full travel.

Q:."' ~s p , u -..: e.W O 3 u 4, J p The control rod testing program was performed on a 24-hour g day, seven-day per week schedule.: However, the test L program was delayed because of difficulties that were experienced with a number of control rod drive units. Initially one of' the control' rod drive units was-in-i advertently operated with the scram discharge valve closed. This action resulted in the unit sticking at the fully inserted position. In response to questioning, Mr. Hess ~j stated that tdue stuck unit could be freed by pushing down j on the top of the control blade. However, this action had to be taken by an individual inside the reactor vessel. Following this freeing action the blade would function at normal withdraw, insert and scram times, but would again stick when returned to the fully inserted position. During subsequent testing, seven additional units with similar maloperating characteristics were found. All defective unit-ere removed and returned to GE for examination,-and new ua.ts were installed. Mr. Hess stated that one of the defective units was operated at the GE-San Jose test stand' and that it, operated normally in all modes except that it also stuck'in the fully inserted position. Mr. Hess also stated thEt GE-San Jose intentionally failed a good control rod drive ' unit by operating with the scram discharge' valve closed. He stated that the scram time increased from a-normal of 1.5 scconds to 6.0 + seconds when the scram dis-charge valve was closed. However, the scram discharge time returned to normal when the scram discharge valve was opened. Mr. Hess stated that an examination of a defective unit revealed that the drive unit inner cylinder deforms when the unit is operated with the scram discharge valve closed. j Following deformation, motion of the inner cylinder is impeded by contact with a drive unit pushing at the fully f inserted position. li In response to questioning concerning the history of opera-tion of the eight defective control rod drive units, Mr. Hess stated that GE knows that one unit was operated with the scram discharge valve closed. He stated that a review of records and logs did not provide any indication

  1. ~

,,,,( 3.- ~~-- -- a..- O O J.; ' / r E of the status of the scram discharge valves for the other A seven units.- Based on the similarity of events, GE concluded that these units were also operated with the scram discharge valves closed by persons unknown. During further discussions Mr. Hess stated that the hand-wheels had been removed from all scram discharge valves 14 early in the testing program and that all units were operated with a torque wrench to prevent excessive forces.from being appl'ed to the valve seats. He also stated that operation i of_the valves can only be controlled by administrative action and that nothing prevents any individual from in-advertently operating these valves. IV. Results of Visits - Bryan This section of the report, which is concerned with the CO review of the facility operating procedures, was prepared by Mr. S. Bryan. A. Scope of compliance Review Operating and administrative procedures were reviewed using CO Manual Draft Chapter 2000 " Facility Procedures" as the basis for evaluation. The scope of the review included a review of the procedure system, emergency procedures, refueling procedure, startup procedures, shutdown procedures, power operation procedures, and 25% of safety related system operating and alarm procedures. Also included in the scope, but not yet i reviewed, are maintenance and chemical control procedures. These procedures had not been completed by the licensee at the i time of the visits. Twenty-five (25) percent of these safety i related procedures will be reviewed when they become available. The radiation control procedures have previously been reviewed by CO.* Followup observations are discussed in paragraph V.E. of this report. B. Oyster Creek Procedure Organization I "The Oyster Creek Nuclear Generating Station Procedures Manual" groups the procedures as follows:

  • CO Report No. 219/68-7, Paragraph II.F.1.

j l 1

._[. 0 3 h, T 100 Administrative Procedures 200 General Plant Operating Procedures 300 Plant Systems Procedures 400 Plant Instrumentation Procedures 500 Plant Abnormalities and Emergency Procedures 600 Surveillance Procedures 700 Maintenance Procedures 800 Chemical Control Procedures 900 Radiation Control Procedures i 1000 Core Calculation Procedures 4 The Administrative Procedures relate to rules controlling personnel behavior within the control room, control of operations p'erformed outside the controi room, procedure violations, procedure changes, logs, standing order instructions, switching and tagging, organization and personnel responsibilities, review committees, visitor control, station operating criteria, test and calibration scheduling, and plant operational philosophy. General Plant Operating Procedures relate to plant opera-tions for conducting startup (normal and scram recovery) from cold or hot conditions, power operation, normal and emergency shutdown from power, and refueling. Plant Systems Procedures are the largest segment of the procedures. They relate to operations for energizing, starting, shutting down, switching to alternate equipment, changing operating modes, filling, draining, and venting. These procedures have sections titled " License Requirements" j in which Technical Specifications relating to the system are listed. They also contain plant operating require-4 ments, precautions, prerequisites, valve checkoff lists l where applicable, and system malfunctions. The malfunction section identifies alarm conditions and specifies corrective action. Plant Instrumentation Procedures relate to operation of nuclear, process, radiation protection, and reactor protection instrumentation. These procedures also contain license and plant requirements for operation.

% __, :12- - r c C 3 / l l Responses to plant alarms and annunciators are listed in Procedure No. 501. The remaining procedures contained in k Section 500, " Plant Abnormalities abd Emergency Procedures" I relate to plant emergencies. 'f Surveillance procedures are formulated to meet test and surveillance requirements of emergency, protective, and standby equipment whose operation is required only during i transients or emergencies. These procedures are a direct outgrowth of the Technical Specifications and are set up to meet surveillance requirements in the specifications. Maintenance procedures are being written for instrument, electrical, and mechanical equipment. In addition to calibration and preventive maintenance activities, these procedures will guide replacement and repair of equipment where failures are anticipated; e.g., control rod drive units and recirculation pump seals. Core evaluation procedures are the procedures used to provide surveillance of the core to insure its integrity during operation. C. Method of Compliance Review Procedures were requested from the licensee noting that they would be returned. A partial draft set was received and reviewed. Much of the effort in this review had to be repeated when final approved procedures were received. Procedures were grouped, and a selection of those to be 2 reviewed was made. The specific procedures reviewed were not revealed to the licensee, unless an item was identified that needed clarification or resolution. The licensee was informed that CO's review was an audit only. l. Administrative Procedures The administrative procedures were reviewed to determine the internal requirements for plant review and approval; the methods for changing, revising, and updating procedures; record retention periods; operators' duties and

~ ~ - y c, ~.L if,*.fR ^ " " ' ' " ~ ~ - " - l ^' p.;. ;.. o. . n;v .o 3 l g Gy * .l S -132.- ] w.... 3 3 ~ N responsibilities; Land identification of the-Technical- -[ i specification administrative ~ controls. These were-judged- } to be adequate. 2. Startup Procedures l e i The startup procedures were reviewed: to determine that ' a - - l: appropriate prerequisites would be performed to make-ready; -l A for' startup, that anticipated critical' rod position would..' l be identified, that rod withdrawal sequence.would be' 1 identified, and that flux monitoring requirements are -l adequate and? identified. In' addition, the review inclu'ded-l a determination that reactivity insertion increments,; .l reactor period limits, and Technical specification'limita-tions'were identified. Deficiencies were identified ~1n 1 ~ the startup procedures and are' listed in paragraph IV.D., -i " Identified Deficiencies". When these deficiencies are I corrected, the procedure will be' satisfactory. ] i 3. Power Operation Procedures l ' 'l Power operation procedure review included' determining thati the licensee had prescribed the reactivity control scheme (when and how to use control rods and-when and how to use. l recirculation flow ' control). 'The3 review also included-l evaluation of the: core flux surveillance, thermal-j hydraulic surveillance, and primary system surveillance., The power operation and core surveillance procedures 1 l reviewed are adequate except for minor deficiencies noted. in paragraph IV.D. l l ~ 4. Shutdown Procedures 'l l 1 Shutdown procedures were reviewed to determine if limita-tions on cooldown were included, if rod-insertion sequences -l were specified, if adequate provisions' were made for decay - - l heat removal, if the flux monitoring requirements were adequate, and if control room personnel requirements were l adequate. The shutdown procedures are deemed to be. j adequate. t a -,... -..,.. -, - - - - ~ -. - ~ ~ - - +. - .,m

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>l 5.- Refuelina Procedures-M ['~ critical multiplication will be adequately monitored, Refueling procedures were reviewed to determine that; sub-l that shutdown margins will be determined, that! fuel' charging' d ^ a increments are specified, and that core alterations will.be. d - predetermined and verified when completed.; Further,.the: jy review included. verification that the Standby l Liquid control- .] 7 i system be required to be operable,.that the refueling equip 0 ~ i ment be tested before s'hutdown,-and'that the_ equipment for 1 each operation be specified. This procedure requires'a i partial rewrite to correct identified deficiencies as ll noted in paragraph IV.D. i t 6. Plant Systems Procedures -l -l Plant systems not related to safety were excluded from the-j list of system. procedures to be-reviewed.. Twenty-five l percent of the remaining. system procedures were reviewed. They were reviewed to determine that-proper steps were j included in the procedure.to permit the equipment 1to perform. l its safety function. This. review,..as well as-'all other j ~, procedure reviews, did not include a step by step evaluation' i of: adequacy, nor'did.it include a verification of the adequacy of identification or correct sequencing'of valve i or circuit breaker check-off lists.- It did include review to determine that provisions -for changing. from one mode of. operation to another. existed and that valve check *off lists' l indicating proper position. er specified operating modes l existed. It also included determining ~that: steps existed' l 1 l for starting.up and shutting.down the system,.that abnormal-- conditions and' corrective actions were ' identified,. and that j effects on the reactor from changing operations were -l -I identified. Only minor deficiencies were encountered.. In j general, ' these procedures were considered to be good. q 7. Plant Instrumentation Procedures t Instrumentation system operating procedures-were reviewed in a manner similar to the system procedure' review. Front panel checks required by the procedures were also reviewed to determine that provisions are made for returning the system to operation. These procedures were judged to be. adequate. '5 -+-n vm.. ~,.mm a c --m me,,~..e m +,w. -. -we -~w ~ < ~ ~ ~ ~ g g y w yw i-r= .w nr w ei--*----- a -e-- m 4~

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dXn 34^-- 7,gs i -J i i '8. Emergency Procedures I i i. All-emergency procedures were reviewed to determine that-r

the procedures -included conditions or symptoms to identify-4 -

i the emergencyi that automatic actions.were: identified; 'j ./ that immediate operator actions:were spelled out, and that' l ,f subsequent actions were included. Unders automatic and-i .4; operator actions, the review was concerned with inclusion 3 of provisions for protecting the. core, i.e., a path to the' I heat sink for stored and decay heat existed or would be instituted; power for emergency systems would be available;-- and improper containment conditions and exhaust system operations exist or will be established. Some' deficiencies i were identified and are noted in paragraph IV.D. t 9. Surveillance Procedures i l Surveillance procedures, as well as all other procedures' reviewed, were reviewed to determine.that operations and: 4 testing requirements were within and according.to the Technical Specifications.. Twenty-five percent of these procedures were-reviewed and found to be; adequate, except for a deficiency.in emergency diesel testing. '(See Paragraph IV.D.) 10. Core Calculation Procedures core calculation procedures werecreviewed to determine that frequency and method of determination'of minimum critical heat. flux, flux distribution, and gross power existed and that the methods.were adequate. The procedures reviewed were' found to be good. D. Deficiencies Identified -i }- The following deficiencies wera identified and discussed with Mr. Hetrick, Plant Operations Supervisor.

? .l Of>wd ^ ^

    • ~"m"

~~~ - '*~% ~ ~-~ ~*L N ? L;g; *: Ub[,,,. h. j3 [ , -7 l l [. et,. + i g ii . Administrative' Procedures (100)- i 4 -1. a. No method provided to effect'a temporary change to. ~.) .1 procedures. j b. Senior operator requirements were omitted. 1 . 3 ' l. y - 2. General' Plant-Operatino Procedures (200) l g

t l

t Approach to Critical and-Pressurization of' Reactor'(203) a. No provisions.-for including anticipated critical ll control rod position in the procedure. -j b. Control rod withdrawal sequence is not provided. c. Control rod withdrawal increment is'. not specified, d. Instrument monitoring guidance and attendance is.not specified. e. Control rod withdrawal pause, after-significant multi-- plication and while.still suberitical, is not required. j f. No guidance on where to level-' power for-heatup. g. Water ' level control point in the_ reactor vessel is not identified'in procedure.* j {- h. Load limiter is to be backed off completely while ) at'less than 10% of rated power. i I Hot Standby (210) .l a. Bypassing a LPRM associated with an APRM is permitted', but rules governing bypassing are not included.

  • See Paragraph IV. F.

___________m.

m. l34lll ~~";.'"~' '~ ~ ' ' ^ ~ ^^ ~0: 1)- .( s., 36 -- J j 4 p ij '. l fo a; 7 4. p- . Power Level Changes (211) I" u-t - A(r-. be recirculation flow adjustment. No provisions are l a. Power! adjustments are permitted by controlirods'and' j ~ h included to prevent reactivity increase:from simultaneous-S operation?of these systems. l y: j b. Inoperable rods are to be valved out. " Inoperable rods"! g,. ~ '. 4 are not defined in the procedure. j .g ' Refuelina Procedures (212) a. Procedure does ' not state that the Standby Liquid Control System must be operable during fuel and control-rod replacement. i P b.. Neutron monitoring during core alterations lis;not specified. e

c. - Shutdown margin-determinations are not,specified until 1

refueling operations are complete d. - No criteria established for use of Standby Liquid Control System or for stopping refueling' based'on' flux'. behavior. 1 1 3.. Plant Systems' Procedures (300) Control Rod Drive System (302) .i a. 302.3.5.2 i I Control rod withdrawal procedure using the notch over-ride is provided for, but no instructions limiting its use are included. l %W u sp .w., l

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v. c h h ?.. ~ c 3 7. - m e' ',e l: i 34 Closed Cooling Water Systems (309) .v .P a a. 309.3.01.3-1 . ( This step warns of equipment heating if the Reactor ~ -l Bui.lding Closed Cooling Water System becomes inoperable. cl It does not list-equipment cooled by this system. rf Fuel Pool System '(311) +. "= j a. 311.3.2 This section of the procedure is for removing the FuelI I Pool Cooling System from service. No instructions;are. x provided to specify when and under what conditions it; .l may be removed, although a statement 'says.'it is. j normally operating. Vital Power System (339y ) a. 339.3.2.1 d This section is for isolating the vital bus.. A reactor-l ' shutdown is required. There_is no provision for first surveying power needs and vital equipment needs before. isolating. Also there are no rules governing removal of vital buses lA2 and 1B2 singly or' simultaneously. .i ' 4. Plant Instrumentation Procedures (400) l a b No deficiencies identified. 1 [ (500) i 5. Plant Abnormalities and Emergency Procedures y 3 Annunciators and Alarms (501) 1 i a a. Corrective action under tha' alarm procedures requires I the operator to " check" a variety of-things depending ~ on the alarm. The " checks" are indefinite and, therefore, ' the instruction is not: clear.*

  • See Paragraph IV.F.

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$v 1, I Station Power Failures (502) s a. 502.2.4.4 I.a Instructions to operator in a loss'of power incident, a after the diesel has started and the' operator is l loading the generator, say "Do not overload the diesela. 7% Yet.there is no statement about. maximum current or. t kilowatts permitted.* .[ b. 502.10.3 On loss of all AC power, procedure has no step to require verification of actuation of isolation condenser. Instrument Air Failure- (503) i Procedure for loss of instrument air has no. reference-a. {- 'to. faulty pressure regulation as the possible'.cause of trouble.- 1 Recirculation System Failure (505) t a. 505.2.c 4-Procedure. incorrectly states that neutron flux will' increase when recirculation flow decreases. t Pipe Rupture (516) t a. 516.4.2.c j A listing of automatic actions resulting from a steam line rupture does not include a statement that a scram + f.. will occur.

  • See Paragraph IV.F.

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__.__a. n. ..s ,'?: G v h-r.' 1 b. 516.4.3.h. I The immediate operator actions following a steam line e rupture should require the operator to scram or verify the scram as the first action. The statement. pertaining 1 l to the scram is indefinite and is included as the eighth 5 item. I l j Significant Increase in Off-Gas Release Rate (517) J ? a. 517.3.a Procedural operator action after a significant increase in off-gas activity requires him to " Determine source of activity increase." This section offers no guidance as to what the source may be or how to search for the i source. There is no reference to clad or fuel failure, nor instructions to check other monitors. Inadvertent Poison Iniection (522) a. 522.1 i The section of the procedure identifying conditions upon inadvertent poison injection from the Standby .j Liquid Control System,.does not include a statement that power will significantly decrease and the tank j liquid level will drop and produce an alarm. t Fire Plan (526) i ( a. The fire plan emergency procedure does not provide I instructions for combating a fire in the control room. 6. Surveillance Procedures (600) a. 601 Auxiliary electric power procedures do not include a functional test to prove that devices which protect the diesels will not be de-energized in a loss of power incident. ~.,;..-

't..,__- ', I ~',". O 3 ! s 1 j 7. Maintenance Procedures (700) i (Not yet received) .L 8. Chemical Control Procedures (800) i (Not yet received) 9. Core Calculation Procedures (1000) k Procedures 1001.1 through 1001.13 have been reviewed and are considered adequate. 10. Other Deficiencies Identified a. Caution notes to warn operators about possible re-activity effects due to xenon or associated with certain operations, such as adjusting recirculation flow and changing feedwater flow, are not adequately interspersed in the procedures.* b. Emergency procedures to include actions to identify and to mitgate the consequences of a fuel failure are missing. c. A procedure to identify circumstances under which the Standby Liquid Control System is to be used is

missing, d.

General instructions should be included to require the-operator to carry out the automatic actions that should take place when plant conditions have exceeded the i protection system setpoint and the automatic action fails.* e. General instructions should be included to require the i operator to reduce power as necessary, or to sustain power losses without return to previous power, when unexplained events happen.*

  • See Paragraph IV.F.

~.....

~ o p, ys_ '.:...a. !T ~ ~ ~ - - ~ - - ~ ~ ^ - " - L, -e. s.. h -) [' 7..,. g j-41 _ A .i-Ji le I E. Resolution of Deficiencies [ The following will be evaluated by Mr. Hetrick to determine if a change in existing procedures is appropriate. f 1. Item h, under 203- + Should the load limiter be backed off completely at 9+l a low power? ~ 2. Ite'> in S01 Should faulty pressure regulation be identified as a possible cause of instrument air failure? Mr. Hetrick indicated that all of the remaining deficiencies identified in paragraph IV.D. will be corrected. 'ollowup Items F. F 1. All of the deficiencies identified above in paragraph D that are not asterisked or discussed in paragraph IV.E.1 and 2, will be corrected prior to the fuel. loading, according to Mr. Hetrick. Their inclusion in the procedures will be sampled by compliance to obtain reasonable. assurance that the licensee has taken appropriate action. 2. The asterisked items will be corrected after startup. when the procedure is revised or rewritten. This is considered acceptable by CO. 3. Outstanding procedures that were not available for CO i review will be written and approved by the plant staff, and a representative sample will be reviewed by CO in accordance with Draft CO Manual Chapter 2000 prior to . f the fuel loading. b ...w ' + - * * - ++-m-pe-w-*==** r v w

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d' ny o 3 lx 4, i e a, 4 V. Results of Visits - Gilbert This section of the report which pertains to CO's followup review in Health Physics, was prepared by Mr. R. Gilbert. For purposes of l comparison, references to previous discussion on the subjects in CO inspection reports is given within the parentheses. j J A. Health Physics Organization & Administration i (219/67-6 and 219/68-1, Paragraph V.A.) l ? I According to Mr. Kaulback, staffing for the Radiation Protection and Chemistry sections has been completed. Messrs. Konta and Heale have been added as assistant technicians.' These two persons had no previous radiation experience. Formal training j to date has included successful completion of the plant indoctrination and qualification courses. Additional training will be given on-the-job by Radiation Protection and Chemistry personnel. Mr. Finfrock. stated that health physics coverage by members of the Radiation Protection Section will be only during the normal Monday - Friday day shift. Coverage on weekends and other -j shifts will be limited to routine surveys by members of the plant operating crew, all of whom will be. considered qualified f by virtue of having satisfactorily completed the station j { qualification course (see paragraph V.B. below. Personnel noting any nonroutine items will report them to the shift foreman 'who in turn will take appropriate action. This might involve notifying the station radiation protection supervisor. As experience is gained, Mr. Finfrock will evaluate this j operating procedure. Should it prove unsatisfactory he has the option of putting health physics personnel on schedule (shift) work or adding more personnel to the health physics staff. CO will followup in this area during the startup test l program to confirm the adequacy of this arrangement. 4 I B. Training Program (219/67-6, Paragraph V.C.4, 219/68-1, I Paragraph V.B.) The basic training course (indoctrination into principles of health physics and radiation protection) has been given to 52 station employees. Attendees, according to Mr. Kaulback, have 4 \\

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. l '.5: 'y& J ,L, j Q "3 -L43 - N [h ~ . i i

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.x ,.e:. been-- persons :with no previous, experience in the nuclear field. occasionally a person with some experience has attended when,- k =< according to Mr. Kaulbackr it appeared in the person's and the t L~ plants.best interest.

- 3 i

l Mr. Kaulback receives the employment recordsL of each person-d i assigned to the plant. Content of the' basic course was' l J reviewed with Mr. Kaulba'ck and it appears to offer station b ! Utt 9. personnel a broad exposure to areas concerned with-radiation 1 } protection. A more detailed' course, in accordance' with statements in FDSAR, Volume I, Section XII.2, relating to on-site training, which is. j a qualification course in station Radiation Protectionf Procedures - has been formulated by Mr. Kaulback. Basically, the course .i content is the material' contained in subsections 901-907 of the Station Procedures Manual. The intent of the course ~is to' provide as-needed instruction to permanently assigned plant personnel so as to enable their being given an " unrestricted l qualification" and accordingly have no or few restrictions' j imposed on access to plant areas. All permanent station personnel must have this qualification. j Daily. tests are given during the 6-day course, and a 6-hour examination following completion of the course. In addition l t to written questions, persons must satisfactorily demonstrate I use of portable radiation monitoring. equipment, perform j surveys, and properly outfit themselves in protective clothing.. About 65 persons have attended the course and all but one-successfully passed. This person subsequently transferred l from the plant. Personnel with previous experience' do not have j to attend the course, but must take the examination. l Mr. Finfrock summarized the present plan for granting contract l personnel access to the plant (unrestricted qualification). l I Before loading starts and restricted areas are established, instruction will be given in sections of the procedures manual applicable to the job requirements. Personnel must then demonstrate an understanding of the station radiation protection procedures before access is granted. l

r V ,f v __.-..m.n...-.m.-- ,p+ i, .) f. } .[ Training prograns for new employees in the future will be 'the same as present programs. Additionally, Mr. Kaulback stated i _f followup and refresher programs will be conducted.- Future plans-for. granting access' to contractor personnel are being j finalized by Mr. Finfrock. i ub The training programs as conducted in the past and planned .} it$ for the future satisfy the requirements of the FDSAR, l Volume I, Section XII.2. Y C. Equipment and Facilities (219/67-6, Paragraphs V.E. and V.Fr 219/68-1, Paragraph V.C.) The status of -portable health physics and counting laboratory f equipment was reviewed with Messrs. Kaulback and Doyle. Both stated that all equipment had been received at the plant. Mr. Kaulback stated that with the exception of two hand and' I foot monitors installed in the monitor and change area, all equipment was locked up in either the emergency control center or the contaminated clothing change room. A summary of this equipment was previously noted on page 46 of CO Report No. ' ~ - 219/67-6, ParagraphLV.F. Mr. Doyle stated that all. equipment scheduled for the counting room had been received, installe.d', and was in operating condition. Mr. Doyle briefly described l [ -this equipment, which is discussed in subsection 909 of the I Procedures Manual. l The inspector toured the monitor and change areas and observed-1 several pieces of equipment, as described by Mr. Kaulback. { The results of discussions with Messrs. Kaulback and Doyle and the tour confirmed that equipment satisfies the statement in the FDSAR, Volume I, paragraphs 6.1.5 and 6.4. The discussion with Mr. Kaulback and the tour also confirmed that protective clothing and equipment needed for fuel loading, normal operations, and emergency conditions has been received-and is in storage. Use of this equipment has been included!in the qualification training course and examinations. (see above). .~,...-.,m .m._ .- - ~. --.. - _ -,._. 5

...j -ge- ~ ~ ~ messe og 3p,^>h h/kmme me a;4w 1 J S =6 nse..: --- c p" 8 :*, 4 0 1 . 4 4 ~ [ - 45:- 6 15 -Both Messrs. Kaulback andlDoyle atated that in:their-g. estimation all facilities required for the' initial. core 1 f-loading'are available for use. Major. equipment is also. N available and operable.. Minor facilities,.such as the 'i byproduct material storage, haveL not been completed and j h probably will not until completion of core loading. This i c i schedule for completion is acceptable to CO.. YJO 1 Discussions with.Mr..Doyle zevealed that GE, as part of the-3 preoperational test procedures, has contracted with Cambridge I I filter to conduct in-place DOP tests of, particulate filters ' and with Barnaby-Chaney to conduct an 'in-place test of the charcoal filters' installed in the standby gas treatment' system. The latter test will use freon gas and be patterned after 'the testing procedure developed at the' Savannah River. plant. The tests will be sufficient to establish the particulate and iodine removal efficiencies specified in FDSAR, Volume I; page V-276 and the Technical Specifications, Section 4.5,.pages-f 4.5-5 ff. Mr. Doyle stated that he has not-as yet developed. station' procedures for' routine filter testing to meet testing' schedules a e s. as specified in FDSAR, Volume:I, pages V-2-8 and 9 and the i Technical Specifications, pages 4.5-5 and 6.. He plans to -l work with the contractors : during the preoperational tests ' and then, based on his observations and the contractor procedures, develop the station procedures.. His statements, verified by Finfrock, indicated adherence to statements made in the Applica-l tion and Technical Specifications. These procedures reportedly ). will be incorporated into section 800 of the procedures manual. {' Mr. Doyle also indicated that analysis of periodic grab samples ~ from the off-gas system and when operating, from the standby gas treatment system via the built-in pressure taps, would 1 provide information as to the integrity and efficiency d the ] filters. ) ) i 2 l

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O ~. .:r ~ 46'- w 1 } [ I 4 e gJ i D. . Personnel Monitorina (219/67-6. Paraaraph V.F.) .y Mr. Kaulback stated that Landauer was selected last year to J provide the monthly film badge service..Use of film badges since that time has been restricted to persons. performing-instrument calibration with radioactive sources. 'No exposures have been received to.date, according to Mr. Kaulback. Both j'k neutron'andh,ibadgeswillbe-used,butthedetermination'of who will be badged with both ' types of. detectors has not been - 't made. Mr. Kaulback is working' with Mr. Finfrock on this matter and expects a decision to be made before core loading. 4 Additionally, personnel criticality monitors have been received, and the assignment of these monitors is being determined. Messrs. Finfrock and Kaulback confirmed that preoperational base line bioassay, consisting of both alpha counting of a urine j specimen and a whole body count, would be performed on all' station employees. Whole body counting will be conducted by 3 Helgeson Nuclear Services, Inc., during the period of February. 17 - 18, 1969. l l ~~ E. Station Procedures Manual - Radiation Protection (219/68-1, Paragraph V.D.) { i l A sampling review of the final approved. version of the station. radiation protection procedures indicates tha' an effort has ) t been made to include more specific information on health I physics operating procedures, counting room equipment and j procedures, and sample counting techniques and procedures. ) Both Messrs. Kaulback and Finfrock are aware that not all i desirable procedures have been included in the manual; however, f they assured the inspector that their evaluation of plant operations has. indicated procedures are not needed for some operations, and for other areas operational experience is necessary before preparing written procedures. This arrange-ment is considered adequate by CO. The frequency of conducting various radiation surveys and the areas requiring such surveys have not been incorporated into the manual. Mr. Kaulback has prepared survey charts and record forms showing such information. These will be maintained

a~ -.... a - .. a y3 M. E... 0 0 4.. f

c in 'the Health Physics office.- The inspector reviewed these

-j forms with Mr. Kaulback and it appears that the survey. program: will be sufficient-to ensure safe working conditions in the plant and effective detection and control over potential' contamination problems. r-F. Startup and Shielding Surveys l l c m.2 j J According to both Messrs. Kaulback and Finfrock,.JC at present l-is satisfied with the startup survey as prepared by GE and ' incorporated into the GE Startup Test-Procedure No. 5 which includes plant surveys. This procedure includes floor plans and data sheets. However, both stated that should preliminary data so indicate, JC would supplement the GE survey program. The JC-GE plans in this subject area are considered adequate by CO. The results will be reviewed on a sampling basis by CO during the startup test. program. G. Environmental Monitoring Stations Environmental monitoring station No. 1 is located at the { meteorological tower (at Environs Monitoring Station No. 3) and was visited by the inspector during the tour. This station is equipped with a Gelman continuous air sampler, rain water l collector device, and area film badge. The station, according to Mr. Kaulback, is similar to 16 others located off-site. l Discussions with Mr. Kaulback indicated that the environmental stations are-located and functioning as specified in Amendment l No. 11, pages I-5-1 and pages following. I H. Radiation Monitoring Systems I i 1. Area Monitors (219/68-1, Paragraph V.F.4.) ~ Discussions with Messrs. Kaulback, Finfrock and Riggle of JC and Mr. Greene of GE, a review of completed pre-operational test procedure D-16, and a tour of several areas of the plant, including the operating floor, established the following regarding this system. All

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%n s. ',.:. : t 1 W :n... K 3 a[ K f - 48 b p fp n q L 32 area monitors (30 originally planned and -2L local GM's ; l k 1.9 J-added to the fuel pool bridge *) have been-installed and I h both electronically'and source calibrated. The source k _ p calibration,'using an 80 uCi co-60 sealed source contained ,y i-in-a GE designed calibration unit '(discussed in GEK-865,-

  • U

' Manual for Calibration), was-made by either GE or by JC 2 personnel under the supervision of Mr.'Riggle. The GE F.,L calibration, procedures, combined with electronic checks c -j-and. previous GE detector calibrations, are' considered by- !~ j CO to provide sufficient information to ensure correct operation of the various units comprising this system. I The techniques for periodic operational tests on area monitors, ' including detector calibration using the 80 uC1 Co-60 source in'the calibrator unit, have been incorporated into section 407 of the station procedures manual. These tests are similar'to the preoperational test procedures, t 2. Process Monitors (219/68-1, Paragraph V.F.1, 2, 3-and 5)

y l

-*( ~ Discussions with Messrs.-Doyle, Riggle and Greene, and a review of'the applicable preoperational test procedures revealed that in all instances the calibrations planned' - {D on the subject systems by-GE consisted of electronic ' and source (sealed) calibration per the vendor recommended 4 I techniques.- For those monitors that. employ-GM detections, -l the detector calibration would :be the same. as that -

i.

J. discussed in paragraph V.H.1 above.. A similar ' calibration unit r reportedly employing a. cesium-137 sealed' source would:be utilized with.the remaining subsystems. In the case'of the liquid monitors, JC plans to supplement the GE-calibration with mockup arrangements to duplicate where possible the actual geometrical configurations involved. Liquid sources with known strength -and concentration would ]. be employed. This calibration would subsequently be verified during actual operation of the facility.- g 1 In those instances where a gas was being monitored,.a. f. series of planchet sources would be made and used by ( JC in conjunction with sealed sources for detector calibration supplemental to that planned by GE.

Again, p'

geometrical conditions would be closely approximated and }: several radionuclides would be employed.

  • CO Report No. 219/68-1, page 49.

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gq y + 'i s. With the-exception.of the-item discussed below, the <4 GE-JC procedures' for, calibration of-the subject; equipment:. o are considered by CO to be adequate. 1 l T . It was noted in the review of_ the GE and JC plans for[ '{ t o J calibration of the' gaseous. monitors (off-gas ~and: stack),. l .1 that no provisions -were included to demonstrate the j m.j adequacy of the isokinetic aspects :of these : systems. J~ In. a subsequent communication with Messrs. McCluskey and. I Hess,'Mr. Carlson indicated'that; isokinetic 1 sampling must' ll be verified. Messrs. McCluskey-and Hess stated.that this' .j ~ would be 'taken into consideration. This matter-remains l unresolved at this time.

I The operational calibration procedures, which will be

.similar.to the preoperational calibration techniques, have been incorporated in the Station Procedures Manual;and I are contained iin subsection 804.. This section'alsol sets-1 r' forth calibration frequencies.which are the'same.as those' j specified-in the Technical Specifications. These calibra-- o tion procedures are considered to be adequate. 1 + ] ~~ 3. Environs Monitors Mr. Kaulbach stated that of'the three environs monitors' "l located on-site, equipment was; installed in only station. No.3 located in the' meteorological; building. 'The_other. i two stations' would. be equipped before core loading. The inspector visited station No..~3Jand noticed that a Gelman. pump and' filter system and'a GM detector andepower supply i -f and recorder had been installed. The filter head and GM i -l tube werellocated in the - roof of the' building. - A totali ~ air flow and total elapsed ~ timer ~ device was. connected to the - 1 j-air pump. -All electronic preoperational. tests as specified j on Preoperational Test Procedure D-18'have been completed-i on this system. Also, many of the electronic tests have' been completed on the equipment-and~ detectors ~for the' i other two sites; however, no detector calibrations have i been made as yet. Mr. Greene stated that the 80 uCi Co-60' .i -l l .-.__.._c ~. y.

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calibration unit would 'be used1to in -place' calibrate the. M GM: detectors.for each-'of the monitors. -(See discussion ((;T

in'H.1 above).

Mr. Kaulbach stated that'JC: Will' review-i the preoperational. test results-.-before.. deciding whether j or.not-to conduct an additional detector calibration. J Scheduled _ operational calibration procedures are presently. fp contained in section-904.10'of.the procedures manual and 19 -indicate use of only the 80 uci'Co-60' calibration unit. 1 Y The : review by the inspector ' indicates this aystem will. be installed according to stipulation ~.in the FDSAR,_ Volume I, page VII-6-7, and that preoperational test procedures are sufficient-to ensure proper operation of the units. J -VI. Exit Interviews -Exit interviews were held with pertinent JC.and GE representatives at the conclusions of some - of the visits discussed in this report. 'In other ' instances, exit interviews were not required due -to.the nature.' of-the ' visit or because the persons involved were in the" company. of the a inspector (s) for most or all'of the visit. -In all cases, the: pertinent. items discussed and the significant comments made by.those interviewed are. contained within the ' body of the report. L _. u. -. - - - - - - ~. -~~~- ~ - - ~~ ~ ^ ~ ~ ~ '

}, r-Q - ~- ^ ~ ^ u;lg,Q%* K !? - ' e -; o. i., l [ TABLE I c,. ]_. t-

u s j '

af STATUS OF PREOPERATIONAL TEST COMPLETION j + Completion Status { Pre-Op Title (%) - 1/27/69 ~.s .C-6 Make-up and Domestic Water System-90 3 . i C-8 . Turbine Building Cooling Water 90 l 4 C-7 {[ C Reactor Building Cooling Water 90 l C-9 Instrument and Service Air-System 100 .[ C-10 Fire Protection System' 10 { C-11 Plant Heating' Boiler 90 i C-12 Vent. Turbine Reactor, Radwaste Access 25 l C-13 Drywell' ventilation 90 l 1 C-14 Standby Gas Treatment System and Reactor 0 .C-16 Liquid Radwaste System 10 C-17 Solid Radwaste System O D-1 Station Grounding -90 D-2 125 D. C. System, Supplement 1-- 100 Battery Capacity Test i D-3 230 KV Electric System 90 D-4 34.5 KV Electric System 90 D-5 4.16 KV Electric System 75 l -D-6 480 V Electric System 75 D-7 220 V/120 V.A.C. 100 - D-8 Emergency Diesel 90 { D-9 Plant Communication System 90 D-10 Feedwater Heater Controls 0 ~ D-ll Feedwater Control 0 D-12 Reactor Protection System 10 D-13 Neutron Monitoring System 90 D-14 T.I.P. Calibration System 90 I D-15-a Procesa Radiation Monitoring 90 -b -c D-16 Area Radiation Monitoring 90 D-17 Off-Gas and Stack Monitoring 50 I D-18 Environs Monitoring (Fixed Stations on Site) 5 D-19 Rod Worth Minimizer 75 I l ? 1 r l n.~ ~ + - - - n-. -= - - ~ ~ - ~ ~ '~-

,,_ y; ~ a--" y '.y w4 1 ego JERSEY CEllTRAL POWER & LIGHT COMPANY 7' 1 d. (CO Report No. 219/69-1) t 't TABLE I L .p STATUS OF PREOPERATIONAL TEST. COMPLETION Completion Status Pre-Op Title (%) - 1/27/69 ,[ A-1 Drywell and Absorption System 0 '*[ A-2 Reactor Vessel Components 100 ,,, ~ A-3 Safety and Relief Valves 0 j A-4 Primary System Expansion 0 1 A-5 Control Rod Drive Hydraulic System '90 A-6 Control Rod Drive 90 A-7 Recirculation System and MG Sets 90 {. A-8 Emergency Condenser 90 Errata and Addendum Sheet, April 5, 1968 A-9 Reactor Shutdown Cooling 90 + A Reactor Clean-Up System 90 A-ll Poison Injection System 90 Supplement #1, special Vibration Test A-12 Core Spray 90 A-13 Containment Spray 90 Supplement #1, Spray Nizzle Flow Tests ~~' A-14 Containment Inerting 90 A-15 Fuel Pool Cooling 90 l A-16 Fuel Handling Equipment 90 A-17 Reactor Head Cooling 0 l B-1 Turbine Oil System 50 B-2 Turbine Control & TG Instrumentation 0 B-3 Steam Cycle: Seal Reg., G1 and Exhaust 0 B-4 Gen. Cooling 0 B-5 Main and Spare Exciter 0 I C-1 Condenser and Auxiliary Hotwell, Vacuum 90 i C-2 Condensate and Feedwater 90 j C-3 C. W. System 90 C-4 Service Water System 90 C-5 condensate Demineralizer System 90 (continued) ._ _,}}

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