IR 05000146/1971003

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IE Insp Rept 50-146/71-03 on 711227-29.Noncompliance Noted: Accidental Releases of Gaseous Activity from Plant Stack in Excess of Specified Limit,Sys Mod W/O Written Safety Evaluation & Failure to Follow Written Emergency Procedure
ML19249A563
Person / Time
Site: Saxton File:GPU Nuclear icon.png
Issue date: 02/09/1972
From: Madsen G, Spessard R
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML19249A562 List:
References
50-146-71-03, 50-146-71-3, NUDOCS 7908230624
Download: ML19249A563 (15)
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Text

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DIVISION.0F COMPI.'IANCE , , , . ' Kt:Giua I - '. - . , CD-Inspection Report No.

50-146/7T-03 Subject: Saxton Nuclear Experimental Corporation- . , ~ . License No. DPR-4 ' Location: 'Saxton, Pennsylvania Priority , , Category C ' . Type of Licenseei E 28 Wt, M . ' ' Type of Inspection: Routine, hmnunced - . Dates of Inspection: December 27 - 29, 1971

. -Dates of Previous Inspection: _ September 7 - 9, 1971 '- . Principal Inspector: Amm[ M 4J > ' ' R. L. Spessa7, Reactor Inspector /' '@ ate d . Accompanying Inspectors: 'Kone . ,.

- Date - . Date ' . Other Accompanying Personnel: None - .. Reviewed By:- , d kri[dm _j [7[n_ - G'. L. Madsen, Reactor Inspector Date " Proprietary Information: None . . ' ' 774063 . - - . . 7508230 ($14 . - .

- ~ ' ._ ~.. ica. - . a.. . - -....-.. - - - -. - -- - - - ' ' ( . . . ' i . _ . . - ,

Section I . .. E'nforcement Action: ~ ' , -A.

Taehn4 cal Spicification I.8 2 - Accidental releases of gaseous ' - act1vity from the plant stack in excess of specified limit.

(Paragraph 12).

. - B.. 10 CFR 50.59 - System-codification without a written safety evaluation.

, .(Paragraph 12) , - C.

Technical Specification N.2.b. (2) - Failure of operating personnel .to follow a written emergency procedure.

(Paragraph 12) . Licensee Action on-Previously Identified Enforcement Matters: ' None reqtiirecL.

i . .j ~ ~ . i Unrecarived Itemst . , The' primary coolant had not been sampled for impurities since returning to operation.

(Paragraph 13) i Status of Previous 1v Reported ITnresolved Items: A.

Auxiliary systems modifications to permit increasing the primary coolant hydrogen concentration have been cc7pleted and appropriate , facility procedures have been reviewed.

However, additional records i are required to complete the QA package.

(Paragraph 14) ! B.

Facility test procedures have. been revised to include acceptance limita or criteria. This ite::t is considered closed.

C.

A change report describing modifications ::mde to the RWDF evaporator feed system was submitted to DRL by letter dated December 9,1971.

This item is considered closed.

Unusual Occurrences: None ' Tersons. Contacted : - , . C.

R~. Montgomery, President, SNEC D. A. Goodman, Superv'.sor, Operations and Testing W.-E. Potts, Supervisor, Reactor Plant Services K. E. Beale, Radiatierr Protection Engineer G. Reid,-Radiochemist R. Walton, Radiochemist: E. Hooper, W On-Site Engineer W4064 '

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. '.. ]- ! ^ , . -Z- ' - ' T. G,6.,w.u, Shfft Suprerviser (SRO)- .- ^ F. Rodies,. Operator (RO)

. P Trexler, Operator (RO) - F. Hertrich, Operator- (RO) ., _ - Management Interviewr . . - . h'he 'following subjects were discussed with Mr. Montgomery, President, ,I ' '5NEC, cfn Yecember 29, 1971: , A., Accidental Releases of Gaseous Activity ' ' ' The.. inspector stated. that the. modification.made to the. caustic addition.line to prevent a recurrence of stem p,acking leakage on valve (7-1231) was made without a written safety evaluation as required by 10 CFR 50.59. After further discussions on 10 CFR , SON sud the importau :e of written safety evaluations, Mr.

Montgomery stated that a written safety evaluation would be prepared i l and' the' modification would be documented in Saxton'~s Monthly Repo,c.

j Mr.. Montgomery further stated that the requirements of 331.1.0 - 1967 Edition would be ma for this modification.

J The inspector stated that a review of the events pertaining to these J.

releases disclosed that operator actions specified in Procedure j' EI-510 had apparently not been adhered to.

Mr. Montgcmery stated

that when. reviewing the events of the December 15, 1971 release, e it should be obvious to anyone that the emergency instruction was ! not followed. He further stated that following the November 29, l 1971 release, EI-510 was placed in the control room and all licensed ' SR0s and R0s were required to read and initial the procedure.

Mr.

- Montgomery stated that following the December 15, 1971 release, he

had persoual discussions with S"EC licensed personnel on this matter.

- He further stated that the complete details of these releases would '

be reviewed by' the SNEC Safety Committee during their next meeting.

. The inspector indicated that enforcement action would probably l be taken on the matters discussed above.

(Paragraph 12) . B.

Auxiliary Systems Modifications o The inspector stated that records indicated an independent audit of the QA package had been performed by GPU and MPR Associates; that facility procedures affected by this modification had been revised; and that these items were considered resolved.

The need for additional records pertaining to three welds in order to complete the QA package was dfscussed.. Mr. Montgc=ery stated ! that Westinghousa would be contacted on this matter, and that the infor=ation would be added to the QA package.

Mr. Montgomery was ' ! l ! 77t1065 !

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. - . 4h 4 that chis matter. was considered to. be. imrmalv.ed. and. that these records vosild be reviewed during the next CO inspection.

!

. (Paragraph 14) , The istspector 1rdicated that the welder's qualification records ~ for procedures used. to weId' austenitic corrosion resisting steel pipe and tubing disclosed that the welders were qualified to the . ] procedures 'in accordance with ASME Boiler and Pressure Vessel Code,.

' - .Section IX, but that the welders had,not been qualified to the ! - proce. dure until after the field welding was completed.

Mr.

Montgomery stated that following the audit performed by GPU and I "MPR, the qualifications of the welders at the time the field ~ welding was performed were questionable, and therefore, it was decided:-to qualify the welderss co the procedures-used. to insure,, that qualified welds could be made. Mr. Montgomery was informed that this matter would receive. further review, at CO:I.

, . ! l In a subsequent telephone conversation on.Ianuary 21,197I, Nr.

Swift, NucIcar Plant Superintendent, was informed that the welder l qualifications for these procedures had been reviewed in CO:I and-were found to be acceptable.

C.

Facility Test Procedures , . ' i' The inspector stated that records indicated testing procedures had

been. revised to includ.e acceptance limits or criteria; that the revised-procedures were being used; and that this matter was , considered resolved.

.- ' ~ D.

Chance Report No. 27 - RWDF Modification , i The inspector stated that the change report had been submitted to DRL and that this matter was considered resolved.

E.

S'amoling of Primary Coolant for Imourities . The inspector stated that chemistry records indicated that the primary coolant had not bean sampled for impurities since resump-l tion of' operations. Mr. Montgomety stated that based on plant i experience. and the midlife fuel examination results for crud i buildup, he was certain the Technical Specifications limit was not being exceeded.

He stated that an analysis would be performed in the near future. The inspector -stated this matter was considered to be unresolved and wenld be reviewed during the next C0 inspection.

(Paragraph 13) ! . - !

! l W4066 i

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r.. - . - F'. Core IIT MfdIffe Cbre FEvsics T'est Results - . ' The. inspector stated that all of the-core physics test results had' not been supplied to SNEC by Westinghouse and, ther2 fore, the hnaetar'.s. review of this. matten was not, complete Mr. Mon?merr --star.ed 'that because-of the, holidays, Westinghousa's revies.of the data ha net been. completed.

He stated that the data should be , . -. completed in about another week and that the data would be commun- ' feated 'to the inspector by telephone.* , , .

G.

Containment' Air Particulate Monitors l The'importance of these monitors and the condition whereby they ,' hava been, frequently out of service for extended periods of time - uere discussed.

Mr. Montgomery stated that this matter would be ' - l reviewed with Mr. Swif t, and that corrective action would' be taken . ! to mini 4ze this condition. The inspector stated this matter would I - be re-inspected during the next CO inspection.

(Paragraph 15) , ' . j - . - . . { - . I'

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  • Data provided in a subsequent telephone conversation on January 14,

' - 1972.

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- q - ., , ! V " . , - . s .. f Section II ' .- N . - Additional Sublects Insnected. Not Identified in Section I. W ere ' No Deficiencies cr Unresolved Items were Found , 1.- General ~ . . J Sinc.e. the lasc Ca inspection, the licensea completed. tha following,

significant' tasks: the auxiliary systems modifications to pereft

I . increasing the primary coolanc hydrosen concentration, reactor - vessel leading, control rod testing, a four hour hot hydrostatic j leak test of the primary coolant syst'em (in service inspection), .-sero power physics testing and startup training (November 10 - , ] 17,1971), power physics testing (November 18 - December 6,1971), - ' AEC license examinations for two cperators (December 7 and-8,1971), 1.

. and full power-operations and load cycling (December 9 - 25, 1971).

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At the time of this inspection, the reactor was in a hot critical - . condition following an unscheduled scram that occurred on December 25,. 1971. The inspector observed reactor startup from this condi- .

tion, load cycle operations, a demonstracian of an in-cure flux-map, charging and letdown operations, RIC-3 response during sampling , . '! 'of the primary coolant, the operation of the stack " Cal" damper, ~! and the surveillance test of the safety injection and recirculation I systems pumps.

There were two unscheduled scrams following completion of core physics testing which resulted from equipment malfuncticns.

i , 2.

Administration and Oreanization a.

Personnel changes b.

On-Site Safety Committee meeting minutes, Septembar 10 through December 29~, 1971 c.

SNEC' Safety Committee seeting minutes, August 31 thro.agh De' cember 29, 1971 3.

Operations , , a.

Future facil!.ty plans-b.

Control Room Log Book, October through December 29, 1971 . c.. Monthly Operating Reports, October through December 29, 1971 d. : Operator performance (observations during the inspection) 4.

Maintenance Records a.

CRDMs and scram breakers (W, Model DB-15) _ . '774068 . -e . ---. .. -

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- ( . . - , Q -.6.- . , . - - . !! - p . w 484 v.McC heeaker, Underground st~rege e-W - cathodic protection syste= j ' e.

' , d. -RWDF preventive maintenance - valves - Experience with limitorque operators e.

f Inspection results of bottom core plate and pilot tubing anti upper core barrel and support plate 5.

Facility P'rocedures . . - . as The following procedures were reviewed: ' (1) OI-418 Boric Acid Removal or Dilution _ (2) OI-454 Crud Filter System , -(3-)' EI-510' High Radioactivity Level (4) S0-8 Ievel and Pressure Limits for Puri'fication Surge Tank (5) S0-16 Minimum Boron Concentration Requirements - , ' b.. The following facility test procedures have been revised ~ to include. acceptance limits or criteria: .(n MI '622. Source calihra. tion. oE the. seeam, generator blow.

- down radiation monitor (RIC-5) (2)' MI 623 Source calibration of stack effluent radiation . monitor (RIC-3) (3) MI 624 24 hour leak test of underground liquid storage . tanks j (4) MI 625 No. 2 turbine overspeed. trip. test' ! (5) MI'621 Control rod drop time test ' . (6).MI' 626 Level test of underground liquid storage tank annu11 (7) MI 5 Reactor plant alternate power supply automatic transfer test (8) OI 413 Nuclear instrumentation (9)~ OI 414 Radiation monitoring. system (107 MI 610' S'af'ety injection system test (1-1) MI 617 Scram circuit response time The following facility procedures have been revised to incor-c.

porate the changes involved in the reactor auxiliary systens modifications: . (1) OI 402a Normal reactor startup from hot shutdown ' (2) OI 403a Reactor startup following scram . (3) OI 407 Main coolant sys. tem cooldown ' (4) OI 408 Filling and. venting of the main coolant system (5) O! 419 Corrosion control agent preparation and addition (6) OI 423 Shutdown cooling (7) OI 438 Instrument and plant air supply, containment vessel (8) 01 440 Shutdown from-pcwer operation (9) OI 417 Boric acid addition to main coolant ' _.

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' . - . ' i . i _7- . - ' ' . _ (10) OI 418 Boric acid removal or dilution - ~(11) EI 509 Loss of main coolant .

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- (12) EI 510 High radioactivity level (13) EI 511 Malfunction of pressurizer power operated re.ef ' and safety valves ~ -(I4)' Er SI3' Tailure of regenitive heat' exchanger (15). EI 515 Malfunction of let. down flow control . ' ' - (16) EI 516 Loss of component cooling - ! - ! . Nate; The re. visions to, the. procedures, (b and c above) were re-l .vievgd by the inspector.

'

. 6.. Feility Surveillance. Test Recuirements, ]1 j The inspector reviewed. the. fc11 cuing ccmpleted test procedures and ! other log sheets for the period September 10 through December. 29, ~ 1971: . % -

a.

Weekly Tests ,

(1) Primary versus secondary heat balance (2) Core flux distribution i (3) Core reactivity ' - - (4) Level tests of the underground liquid storage tank annuli ' i (3 times / week) . I '

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b.

Monthly Tests t I (1) Calibration of the failed fuel element detector e (2)- Calibration of the steam generator blowdown radiation monitor (3) Calibration of the stack effluent radiation monitor (4). Calibration of the radiation monitoring system (5) Safety injection and recirculation pumps and automatic startup control (observed the December 29, 1971 test) ' c.

Semi-annual Tests (1) Control rod drop time (2) Scram circuit response . (3) No. 2 turbine overspeed trip (4) 24 hour leak test of underground liquid storage tanks 7.

Primarv System ' a.

Primary coolant chemistry logs (November - December 1971) for , the following parameters which have Technical Specification limits: . (1) Chlorides riri4070 __

_.. _. ... - ' ' ~ .m p . [ - 'd - .._ /

t. k - ' -a- . - . ! . (fi oxygerr _ . . . - (3) Boron . ~ (4) Lithium.

~ - - (5) Hydrogen - - - (6) Rnd-toeet-ivity . . i br.

Make-up water to reactor plant chemistry logs- (November - ' j . ' December.1971) for the following parameters which have Technical ' .1 - Specification limits: i . (1) Conductivity . .- (2): Chlorides ~ .~ ~ (37: Sfli'cotr dioxide . Hydrogen addition control, procedures and records (November - ac.

December 1971) - d.

Four hour hot hydrostatic test (in service inspection), completed test procedure, dated November 8,1971.

8.

Reactivity Control and Core Physics a; 1% shutdown margin requirements-I

b.

Core power distribution limits (data from December 13, 1971

flux map) for the following: I, - (1) Loose lattice assemblies, peak pellet (2) Load follow assemblies, peak pellet (3) PNC assembly at peripheral core position D-5, peak pellet I (4) Test fuel rods for load follow assemblies (TS Change 44), ! peak pelle.t - i (5) Loose lattice rods surrounding solid rirc-4 rod (TS Change 47), peak pellet (6) PNC test fuel rods for loose lattice assemblies (TS Change 41), peak pellet (7) Power transient test fuel rods for peripheral subassembly at core position N-3 (TS Change 48), peak pellet (8-) Plutonium oxide fuel rods for central subassembly (core position N-1), peak pellet.

- c.

Midlife core physics test results for the following parameters'. Core Parameter Core Power (1) Mod. Temp. Coeff.

ARO (All rods out) HZP (Hot Zero Power) ! "I'/4071 - - ._.

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p..,..+....- - - _ ' ~ r md . ' . ' ,- . - . . . ~ a - - -9- ' . '- - .. ~' ' R2 In.(Centrol rod No. 2) HZP j '"ARO-20 MW ' . R2 In .. 23 MW.

- . . (2) Power Coeff.

- 10-20 MW . 20-25* MW .- -- - , _ >;. . . _ d ~ (3-) Boron Worth EZP j - - - 20 MW.

. ,, ._ - - . .- (.41 Peak Xenon Worth - 23 MW ' ~ 6 E'rs,. . , , - - after shut ~down ... . . (5) Pressure Coeff.

HZP ~ - - . . , . (6), (a) Rod. 2 Worth

  • HZP

- 10.MWt - (b) Rod' 5 Worth HZP ~ 9.

Auxiliar Systems l ' Regenas ative heat exchanger - the QA package for the relocation of the safety valve, V-53, was reviewed to insure that the safety . considerations described. in the licensee's Change Report No. 26, ~~ '

dated December 9,1971, had been met.

I '10.

Containment . l ' Eydrostatic test records of two contairment vessel (CV) penetra-tions. used for the pipe that connects PSV-501 (purification surge ! tank relief valve) to the CV discharge tank (October 28, 1971).

l Note: This line was rerouted and therefore both penetrations were required to be tested by the Technical Specifications.

11.

Radiation Protection a.

Radioactivity samp' ling records of the.refue!.ing water storage tank, September through Dece:her 1971 b Personnel exposure records, August through-Cctober 1971 c.

Observation of a radiation survey of the primry system let- , down line performed by one of the shif t operators d.

Gaseous 'and liquid release permits, September through November 1971 Control of gaseous releases during primary coolant sa=pling e.

and operation of the charging pumps (required be.cause of high gaseous activity, mostly xenons, in the primary coolant) t .

  • Extrapolated from 23 MWt; 25 MWt projected.100% power.

774072

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- , ". . Datails of Subjects Discussed in Section I ] . . . ~ 12. Accidental Releases of Gaseous Activity from the Plant Stack . References: Inquiry Report Nos. 50-146/71-02 and 03 and

,_ Licensee letters-to DRL, dated December 9 & 28, 1971 - ,

i Additional information relating to the~se releases which was obtained.

-i . .during discussions with Messrs. Montgomery and Potts and from a '

. - review of site records, procedures and equipment is suz::merized as 'follows.

, I a. November 29, 1971 releases - There were three separate releases . from the same source (stem packing leakage from valve V-1231 in a caustic addition line which is connected to the purification , sy e am letdown piping in tua charging pump room) during a se.ven hour period.

The release concentrations, excluding I-131 and averagedovera15minugeperiod,forthesereleaseswere - 3.52 x 10-3, 0.81 x 10- and 1.08 x 10-3 uC1/cc respectively.

The release-concentrations, averaged over a.15 minute-period, for the first and third releases were in excess of the Technical ' Specification limit of 1 x 10-3 uCi/cc.

(Paragraph N.7.b) Paragraph N.2.b.(2) of the Technical. Specifications specifies, in part, that Standing Orders to operating. pe::ennel shall !- require that written procedures and instructions provided for emergency conditious shall be followed in conducting activities identified therein.

SO No. 18 is used for promulgating these ' requirements.

Emergency instruction for High Radioactivity ! Level (EI-510) requirea,. in part, that when the stack gas (RIC-3) alarm is actuated, i==ediate action consisting of the j following will be taken: I l (1-) Attempt to reset the-alarm and check. the insert: ment for-proper operation and verify the alarm with a portable monitor or air sa'nple.

(2) Temporarily place the fresh. air damper to the " Cal" position.

- If the counts do not. decrease, this-is evidence of increased ! - background.

If the counts do decrease, this is evidence of - a radioactivity release and the following action is required.

(3) Terminate or reduce the rate. of activity release resulting from the following:.... (6) Eoron dilution.... (9) Pressurizer vent line operation.

- - 774073 . m _.....__m..- _.m- . _. - - . _ _. _ _ _

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- n- - - . . . . . - _ ! ' . During the first release and after RIC-3 had been verified to be

operating correctly (by an instrument technician), a period of. about

10t minutes elapsed before action was taken to reduce the rate of

. ', activity release, i.e., boron dilution terminated by operator action from the contrcrl roo::r. RIC-3-indicai:ed a release concearratiotr of ' greater than.1 x.10-3 uC1/cc for a period of about. three minutes.

. c prior to isolation and a-concentration of 1.17 x 10- uC1/cc at

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' ' the. ~ time of isolation.

. . . . The ten minute period which passed before action was taken to.

.) - ' :.- tambte the release does not appear to meet the i= mediate j - . j . action requirements specified in EI-510.

I The licensee's investigation to identify the source of leakage and the corrective actions taken (tightening of valve. packing on suspected valves in the charging pump room and operating -

alternate charging pumps for charging and letdown) prior to the d ' ' third release indicated that the source of leakage was some where j in the bleed line downstream of valve (V-114). The third release n occurred' when bleed and feed-operations were resu=ed (V -114'

opened):. The source of leakage was identified on the following dayOe pressure testing the bleed line (from valve V-114 to the , ~RWDF storage tank) with nitrogen at 100 psig, and the leakage

j yas stopped by tightening the packing nut on valve V-1231.

A review of the events disclosed that the packing nuts on valves V-1231 and V-1237 were tightened af ter the second release and not af ter the first release as previously reported by Mr.

Montgomery during a telephone conversation on December 1,1971 (C0 Inquiry Report No. 50-146/71-02).

i The inspector asked Mr. Montgomery why bleed and feed operations , i were resumed following the second release ~ since the source of j leakage had been isolated to this line and equipment for pressure

tasting. this. line with. nitragan was, readily availabla for use-Mr. Montgomery stated that bleed and feed operations were re-sumed to see if the source leakage in the line-could be identified and to check whether-or-not the corrective actions: taken (tightening of valve packing on valves. connected to this line) . had been effective.

The inspector stated that it would seem more prudent to perform this check with nitrogen rather than with zenon gases.

, The correctiva action taken by the. licensee consisted of the ! following: (1) Operate with globe valve V.-1231 open and back seated so the stem packing is not pressurized by the letdown flow.

(2) A second globe valve, V-2461, was. installed in the caustic addition line outside of V-1231 to provide isolation capabilities.

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. - , . l - 12 - .. , . - _ (3) The packing of all valves and all swagelock fittinsts in

the charging pump room have been tightened.

The stem packings . ' and bonnet gaskets of about 25 valves in the underground ' - ' storage tank piping have been leak tested and the remaining ' valves in these systems are sche-duled to be lenk rested as ~~ - - operations permit.

. , t .. .. , i l - ,(4) The licensee's, letter to DRL, dated December 9,1971 with a- ' - cover letter of'the corrective actions taken was sent to d-all members of the SNEC Safet'y Committee.

i - - .. . When the inspector asked Mr. Montgomery to show him the written - '_ safety valuation covering the installation of valve V-2461, g Mr. Montgomery stated that a written evaluation had not been prepared. He stated that this change had been discussed among supervisory personnel and it was concluded that this change ~ would be a prudent one . I bL.

December 15, 1971 release - Moments af ter the vent line was

puc..in service, RIC'-3 began to increase and within fiv'e minutes - its alarm point was reached.

Then,. a period of about 21 minutes elapsed before the. fresh air damper was placed to the " Cal" position to verify the release and about one minute later the ~ vent line was isolated by operator action from the control room. RIC-3 indicated a release concentration of greater than i 1 x 10-3 uCi/cc for a period of about 14 minstes prior to opera-i tien of the fresh air damper and a concentration of approximately ' ' - 1.22 x 10-3 uCi/cc when the damper was operated.

During the . release and prior to operation of the fresh air damper, radiation surveys were made of the plant including charging room, auxiliary equipment room, yard area (includes RIC-3) and RWDF. The surveys revealed an increase in background activity; however, Saxton ' ' personnel did not interpret the survey results to be indicative . of a release.

' The 22 minute period which passed before action was taken to ' verify the release, as indicated by RIC-3, and reduce the rate of activity release does not appear to. meet the i==ediate action requirements specified in EI-510.

. The pressurizer vent line isolation valve, PC-97V, is a. Mason Neilan, forged stainless steel 3/8" globe valve, rated at 2485 psig at 6500F. The valve packing (source of leakage) is John Crane, Type 2CR-J (asbestos graphite) with 10-11 packing rings and is recommended for liquid and gas service.

The pressurizer was being vented for the first time since returning to power on .cember 9, 1971.- Prior to December 9, it. had been vented intermittently for about 20 hours without incident.

Mr.

Montgomery stated that this operating period (venting prior to _.. __ ._ _ 774075

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. - 4.c . -. . - 13 - ~ ' - ~ . - . ~ . the release) had been mistakenly reported in SNEC's December 28, - C ~1971 letter as having occurred during the period December 8-15.

" ' .Mr. Montgomery stated that the valve manufacturer had been contacted,

and it, was recommended that the valve be cycled several times and. its, stem packing tightened following each cycle until the ' , ",. packing' was tight. He furthen stated that 'a hydrostatic test.

- pmcefure. to insuna leak tightness, of' the sten packfug would, be.

_ .- prepared and reviewed by the safety committees prior to use.

. , ~ Acc.ording to Mr. Moo *gomery, a pressure test with. nitrogen - ',. - up to 2000 psig wouJ. be possible. The vent line will remain - - ~ isolated until PC-97 ' is satisfactorily tested.

1A Sampling of Primary Coolaut for Impurities - The Technical Specifications (paragraph N.4.b.(6)) specifies that for. power operation above 1 MWe, impurities in the primary coolant will be less than 5 ppm. During a review of chemistry records, the inspector noted that the primary coolant had not been sampled for impurities since returning to operations following the recent ex-tended' ontage.

. Itr separate discussions witir Messrs. Reed and Potts, the inspector was-informed that with the high gaseous radioactivity levels in the . primary coolant, the Plant Superintendent,. Mr. Swif t, would not allow this sample to be taken using the former procedure because of the large volumes involved. The maximum valve (impurities) measured- . . in the past was reported to be about 0.2 ppm.

The crud buildup '

. of fuel rods found during the midlife fuel inspection were reported by Mr. Potts to be minimal.

Mr. Potts stated that there were plans to pull a sample in the near future using the newly installed crud sampling systecr.

. 14. QA Package - Aux 111arv Systems Modifications

References: Licensee's Change Report No. 25, dated October 18, 1971 and Amendment No. 1, dated November 10, 1971, submitted to DRL , . A review of QA records and discussions with Messrs. Goodman, Potts ,

and Montgomery disclosed that the safety considerations described , in paragraphs 3.1 through 3.6 of Change Report No. 25 had been met; however, the. inspector identified three welds which required additional supporting records to complete the QA package.

These were:

Shop-veld on valve V-2456 (pressuriser vent line) - The.only a.

documentation available was a referenced Westinghouse job order number in the. bill of materials, - 774076 . _ -. _... -...

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h.. Weld.Noc. 44 and 51 (boric acid system) teere identified en the a

weld. inspection record as being repaired. The weld inspection record also showed that a satisfactory LP test of these welds had been performed.

Tbe welding procedure specified that weld defects and method of repair would be documented.

' ~[ InJa subset {uent telephone conversation with, Mr. Swif t, Nuclear J - PTant Superintend ~ent on January 2I,1972', the itspe'etor v:rs informed' $ hat' the additional records..for these-welds were available. and would ' ' .be added-to the QA. pa+P.. . .- t _ - ,. . - - ' 15. Containment Air Particulate Monitors.

The-containment vessel air is monitored for particulate activity by RIC.1 and 11 and for gaseous activity by RIC-2.

The licensee's ^ ~ letter to DRL, dated October 5,1971, describes methods for identi-fying leakage from the primary coolant system which includes detection by these monitors. The licensee considers, as stated in the letter, the particulate monitors to be more sensitive in leak detection than the-gas monitor.

. . A. review of equipment trouble reports for the period.of September through December 28, 1971 revealed that RIC 1 and 11 experienced filter feed failurcs on a weekly basis-during this period, Itr discussing.this matter with Mr Potts and reactor operator personnel, the inspec. tor was informed. that this. condition results when. the filter paper runs out; that this frequently occurs on a back shift , - or over weekends; and that this results in the monitors being out of service until the next day or Monday morning because the paper is changed by instrument technicians. Mr. Potts stated that during periods when the reactor is shutdown and maintenance is being per-i formed in the containment vessel, these monitors are always kept g operational.

He also stated that he believed RIC-2 provided adequate i backup during operation because of the high gaseous activity. in the.

primary coolant.

Saxton's Technical Specifications.do not contain a minimum condition for operation for containment vessel air monitors; however, the Technical Specifications do require that these monitors be operational j-during maintenance activities performed inside the containment vessel.

. . . 0. I W4077 ._ -._ .__

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. ' ~ U. S.. ATOMIC ENERGY COMMISSION ,

REGION I

. I . DIVISION OF COMPLIANCE l Report of Inspection

! CO Report No. 146/70-2 Licensee: SAXTON NUuM EXPERIMENTAL CORPORATION License No. DPR -4 I Category C Dates of Inspection: September 1 - 3 and 23,. 1970 Dates of Previous Inspection: April.20 - 23, 1970 Inspected by: e

0 R. J. McDermott, Reactor Inspector (Responsible) Da'te . ___ hl O R. L. Spessar8, Reactor Inspector (Wrote Report) Date Reviewed by : a h io O R. T. Carlson, Senior Reactor Inspector 'Da'te Proprietary Information: None SCOPE A routine, announced visit was made'to the Saxton Nuclear Experimental Corporation (SNEC), Saxton, Penncylvania, on September 1 - 3, 1970 to tispect the 28 Mwt pressurized water reactor and for official turnover to the newly assigned principal reactor inspector, R. L. Spessard,by R. J. McDemott.

A subsequent corporate level interview was held at the Saxten facility on Septembar 23, 1970 and was directed toward Provisional Instruction 1800/2, Chapter 1830, " Corporate Level Interviews" and significant observations made during the September 1 - 3, 1970 inspection.

SLTfARY ' Safety Items - No safety items were identified during this inspection.

Non:empliance Items - The following three items of noncompliance were noted, and a form AEC-592 has been issued.

1.

Paragraph N.1.a. (2)(c) of the Technical Specifications requires, in part, that minirm.:m qualifications consisting of five years' experience in ' engineering, operations, and maintenance at nuclear or fossil-fuel power plants or similar facilities, with two years in a responsible super-visory position of such facilities, and qualification as a licensed senicr reactor operator, shall be maintained for personnal occupying the position of Supervisor - Reactor Plant Services.

7"74078 908280 h Y

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2- - Contrary to this requirement, the minimum qualifications with regard to experience and licensing were not maintained when Mr. W. E. Potts assumed - the responsibilities of Supervisor - Reactor Plant Services on August 28, 1970 following the departure of Mr. J. G. Herbein.

- ., 2.

Paragraphs N.l.a.(1) and (2) of the Technical Specifications set forth the organization for the project and for the conduct of plant operations , ' in Figures N.1.a.(1) and (2) which include, in part, test engineer

position (s).* Contrary to these requirements the test engineer position (s) are vacant.

3.. Paragraph I.8.d. of the Technical Specifications requires, in part, that the radioactive concentration of gaseous releases as measured by the ' radiation monitor in the ventilation duct ahead of the stack fan shall not exceed an instantaneous concentration of 1 x 10-3 uCi/ce, excluding I-131, ! when averaged over a 15 minute period.

' ! Contrary to this requirement, the stack monitor indicated a radioactive , concentration in excess of 1 x 10-3 uCi/cc for a period of 3 - 4 hours during an accidental gaseous release occurring on May 14, 1970.

Unusual Occurrences - i. Section E.1. of this report contains information concerning the separating , of the sensing lines to the pressurizer level indicator D/P cell during primary system heatup.

2.

Section I.2. of this report contains information concerning the shearing of the 20 ton bridge crhne cable.

3.

Section O. of this report contains infomation concerning a dropped irradiated fuel subassembly.

4.

Section Q.4. of this report contains information concerning two unplanned releases of gaseous activity to the environs.

Status of Previously Reported Problems - None Other Significant Items - 1.

An IBEW strike involving Saxton's operators and technicians occurred during the period May 27 - June 20, 1970.

(Section C.)

2.

The licensee has conducted an emergency preparedness drill since the last inspection. (Section D.l.)

, 3.

Saxton's primary coolant activity reached its highest level, 180 uCL/ce, just prior to shutdown for fuel inspection.

(Section E.2.)

774073 ,

  • Also, Section 301.B.2 of the Final Safeguards Report specifies two test engineer positions.

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. -3- . 4.

Fuel pin inspection was conducted at Saxton by the Westinghouse Mobile Fuel Evaluation Team and fuel pin failures were identified.

(Section G) 5.

Saxton's containment bridge 20 ton crane has been inspectad and load tested since the last CO inspection.

(Section I.2.)

6.

The licensee reported errors made in calculating Noble gas activity l concentrations and has corrected his procedures and gaseous releases j for calendar year 1970.

(Section Q.3.)

,

Management Interview - September 3,1970 - Messrs. Swift and Goodman represented SNEC during the management interview held at the conclusion of the inspection. CO was represented by Messrs. McDermott and Spessard.

1.

The basis for not correcting past liquid radioactive waste releases due l to calcul!ation errors was discussed.

The inspectors pointed out that there j . were no records to indicate that liquid radioactive waste was analyzed for ! dissolved gaseous activity prior to release. Normal procedures call for

boil down of samples prior to analysis.

Mr. Swift explained that a substantial portion of the dissolved gases is evolved during the normal ,- processing of liquid radioactive waste and this portion is collected in gas - decay tanks.

However, he stated that an analysis would be made to

subscantiate the basis for not correcting past liquid radioactive waste releases.

.

The recent significant organization changes at the Saxton facility were discussed. The inspectors pointed out that the qualifications of Mr. Potts, Supervisor - Reactor Plant Services, did not meet the minimum. requirements - for this position as set forth in the Saxton Technical Specifications, and that the vacancies La the test engineer positions did not meet the project organization requirements as set forth in the Saxton Technical Specifications.

Mr. Swift stated that these changes had been discussed with DRL during a May 6,1970 meeting and that DRL indicated that although the qualifications of Mr. Potts were marginal, he was acceptable.

Mr. Swift further indicated that recruitment for the test engineer positions was in progress but had no idea when these positions wculd be filled. The inspectors informed Mr. Swift that further compliance action would ' probably be forthcoming.

3.

The dropped fuel subassembly occurrence on August 6, 1970 was discussed.

The inspectors pointed out that their review of the On-Site Safety Comrittte minutes (last meeting en August 20, 1970) indicated that this occut.ence was not reviewed by the committee and that according to the Saxton Technical Specifications, plant operations to detect potential safety hazards are to be reviewed by the coesittee.

Mr. Swift stated that the committes had not reviewed this occurrenca, but would do so at the next scheduled meeting. The inspectori expressed their concern at the apparent lack of attention given to this potentially hazardoms occurrence by Saxton management personnel.

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4.

The unscheduled releases of gaseous activity to the environs wnich occurred on May 14, 1970 and August 26, 1970 were discussed. The inspectors i pointed out that records indicated that during the releases the stack monitor (RIC-3) was pegged full scale for a period of 3 - 4 hours for the ' - May release and a period of about 5 minutes for the August release.

The inspectors further stated that the calibration curve for RIC-3 indicated g that when RIC-3 is pegged ( p 1000 eps) the Saxton Technical Specification lbnit for an instantaneous concentration of 1 x 10-J uCi/ce, excluding

i I-131, when averaged over a 15 minute period is being exceeded.

The

inspectors pointed out their reservations concerning the ability of the i stack monitor to detect accurately gaseous activity concentrations in the range of 10-3 uCi/cc. and higher due to its limited range and therefore had reservations as to exactly what concentration was actually released.

Mr. Swift stated that the calibration curve for RIC-3 presently in use was the original calibration curve made in 1962 and that subsequent calibration checks indicated that the original calibration was conservative and there-fore was still being used.

Mr. Swift also pointed out that during the May

release gas samples taken at the stack indicated that the activity was predominantly Xe-133 and the concentration was 1.2 x 10-5 uCi/ce.* Mr. Swift stated that the calibration of the stack monitor would be varified, but if the calibration was found to be conservative, it would probab1v be left as is.

It was the inspectors' position that an accurate calibration was necessary and that utilizing a monitor with a higher ' response capability should be considered.

Mr. Swift was informed that , further compliance action would probably be forthcoming.

5.

The condition of the carbon steel piping and valves in the Radioactive . Waste Disposal Facility (RWDF) was discussed. The inspectors expressed their concern about the present integrity of both the gaseous 'and liquid waste disposal systems in light of the two recent unscheduled releases of gasects activity.

Mr. Swift stated that an active maintenance program is in pr sgre. s on the gaseous vaste disposal system which consisted of inspection, repair, and testing. He stated that this system was about 50% complete and that valve parts had been ordered. The inspectors indicated that the progress of the maintenance efforts would be followed in future inspections and that they would look for a program of scheduled preventive maintenance on both the gaseous and liquid waste disposal systems.

Mr. Swift stated a program would be developed and implemented by the next inspection.

6.

The recently identified fuel pin failure was discussed.

The inspectors asked Mr. Swift to consider the reportability of this information to DRL, since as a licensing body DRL was interested in fuel cladding performance of ths various types of fuels they had approved for use in the reactor.

Mr. Swift stated that this matter would be considered.

7.

The significance of C-14 activity based on studies by the Public Health Service was discussed.

The inspectors asked if the licensee would consider attempting to identify the presence of C-14 activity.

Mr. Swift stated that an analysis to determine the presence of C-14 activity would be censidered.

  • Inspactors' Note - Sample taken late in release period.

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- . . -5-l Manage _ ment Interview - September 23. 1970 - Messrs. Montgomery and Swift represented SNEC during the interview. Messrs. Goodman, Potts, Beale, Reid, and Pekar fromj SNEC were also in attendance. CO was represented by Messrs. Carlsen, McDermott and Spessard.

itt. Carlson stated that the, purpose of the meeting was twc-fold: (1) to meet the ' i CO progran objective of holding routine, periodic get-togethers on the corporate " level to review matters of mutual interest including, as appropriate, an updating in the areas of regulatory philosophy, practice and organization, and any necessary j clarification of the CO role and inspection program; and (2) to discuss the results of the last inspection.

t l 1.

Mr. Carlson reviewed the current organization and functions of the Regulatory, with emphasis on DRL and CO, with respect to the various - roles performed during tha life of a typical facility starting with the initial submittal of an application for a construction pernit through , i normal co=cercial operation.

Special attention was given to the changes

made in the CO organization and program. Areas covered included: ! type, frequency, scope, and depth of CO inspections; inspection techniques; ' basis for inspections; inspector qualifications; use of consultants;

methods and types of enforcement actions; management meetings; emphasis on QA/QC and the applicability of proposed Appendix B of 10 CFR 50 to I operating plants; and the requirenents and intent of 10 CFR 50.59.

It f was emphasized that CO inspections are performed on a sampling basis and do not replace the need for the performance cf comprehensive audits of operations by licensee management.

The requirement that the licausee be able to demonstrate complianca with applicable regulations and the related C0 need for accers to information were discussed.

The philosophy , behind typical incident reporting requirements including the concern as to the possible applicability of a particular probics to other facilities, was emphasized.

2.

Mr. Carlson stated that the results frem the September 1 - 3, 1970 inspection vere generally satisfactory although it may not always be apparent since inspectors tend to be problem oriented. He noted that the following observations ware made.

, . 'a.

Several itc.- of a followup nature had been satisfactorily accomplished by the licensee.

(Sections D.1., I.2., and Q.3.)

b.

Several new items had been identified and discussed at the September 3 exit interview. These itams (Paragraphs 1, 3, 5, 6 and 7 of Exit Interview and Section I.1.) were again discussed individually nith the licensee during this meeting to ensure a common understr.ading of the proposed resolutions. The licensee's proposed resolvuions remained as previously stated to the inspectors.

Mr. Swift reported that inspection and repair of the piping anc valves in the gaseous waste dispossi facility have been completed and that inspection and repair of the piping and valves in the liquid waste disposal system have been initiated.

(Paragraph 5 of Exit Interview).

Mr. Montgomery reported that all information available to h0n concerning the inspection of . W4082 . _.

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. - . the failed fuel pins had been reported to DRL during subsequent tele-phone conversations and that this information would appear in the, s semi-annual report to DRL as - quired by License DFR-4.

He further ' stated that he did not know whether or not Westinghouse would subait a detailed report of the inspection to DRL because of the proprietary information involved.

Mr. Carlson stated that DRL understards the nature of proprietary information and handlas it quite often.

' Mr. Montgomery requested the aid of compliance La obtaining additional information concerning the measurement of C-14. (Subsequeatly ,commnnicated to Mr. Swift by Mr. Spessard via telecon.)

l' - c.

The two items of noncompliance pertaining to plant organization were discussed at length both during and subsequent to this meeting.* ' Factored in were the results of the discussions held on this subject ' (plant organization) during the May 6,1970 DRL-Saxton meeting, including the mechanism by which these issues would be resolved by the licensee. -In summary, the licensee (Mr. Montgomery) has agreed that formal resolution of thesa issues is in order, and proposes to accomplish it as follows: (1) Test Engineers - To submit a proposed change to the Technical Specifications to allow filling of these positions with already present Westinghouse employees.

I (2) Qualifications of Mr. Potts - To document in a letter to DRL the ! substance of tL. day 6 discussions, including their justification for assigning Mr. Potts to the position of Supervisor - Reactor Plant Services, their proposed short and long range plans to up-grade his qualifications, and their plans includ'-tg *.imetable to have Mr. Potts become a licensed senior reactor operator.

- d.

The item of noncompid ance pertaining to the May 14, 1970 accidental gaseous release was also discussed. The licensee (Messrs. Montgomery and Swift) stated that the following actions had been taken: (1) Ir. proved the integrity of the gaseous vaste disposal systen by 'nspection and repair of 2007. of the piping and valves in the system.

. (2) Reviewed the venting procedures for the surge tank and considered possible isolation procedures.

(3) Routine pressure testing of the vacuum regulating valves and the gas compressor system is being performed.

(4) Reviewed possibilities of expanding the range of the stack monitor; however, this doesn't appear to be feasible.

It was Compliance's position that action (4) above r'ould be pursued further because the licensee could not adequately measure a burst type gaseous release.

Mr. Montgomery stated this would be reviewed further.

Mr. Montgamery was advised that a, form AEC-592 would be issued regarding the thras items of noncompliance discussed.

7gg

  • Fost-meeting discussions included telecons between CO (Mr. Carlson) and DRL (Messrs. Schemel and Woodruff) and DRL and Saxton (Mr. Montgomery) on 9/23 and 24/70

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' . . ' , 7- - . DETAILS A.

Persons Contacted: - SNEC , C. R. Montgomery, President (Corporate Level Interview Only) R W. Swift, Nuclear Plant Superintendent . i D. A. Goodman, Supervisor of Operations and Testing l W. E. Potts, Acting Supervisor of. Reactor Plant Services - ! K. E. Beale, Radiation Proteccion Engineer G. Reid, Rat.iochemist i j S. Pekar, Maintenance Foreman

~ l B.

Administration and Organization J 1.

Training . Two senior reactor operators have been licensed at the Saxton facility since the last inspection. Mr. Swift informed the inspectors that one reactor operator and one senior reactor operator are currently in training and will take license-examinations this fall. Westinghouse presently has , 10 customer personnel in training at Saxton.

I 2.

Personnel Changes . . Mr. Swift informed the inspectors of the following significant parsonnel changes:* a.

Mr. E. A. Liden, Supervisor of Reactor Plant Services left on May 22, 1970, and was replaced by Mr. J. G. Herbein, Supervisor of Oper.tions and Testing.

Mr. D. A. Goodman, Test Engineer, vacated his position to assume the responsibilities of sunervisor of operations and testing on the same u?te.

b.

Mr. J. G. Herbein, Supervisor of Reactor Plant Services, left on August 28, 1970, and was replaced by Mr. W. E. Potts, Test Engineer and a senior reactor operator in training.

Mr. Potts spent six years in the U. S. Navy, the last two of which were on the Nuclaar Submarine George Washington in the Inertial Navigation Group where he supervised 12 - 15 personnel. This was a non-nuclear duty position.

He worked for three months as a student engineer at Saxton during the su=mer of 1969 and after receiving a BS degree in e'.ectrical engineering from Pennsylvania State University in March of 1970, he returned to Saxton and assumed the responsibilities of a test engineer.

The qualifications of Mr. Potts are not in accordance with the.minir2m requirements for the position of supervisor of reactor plant services as set forth in Section N.1.a.(2)(c) of the Saxton Technical Specifications which requires in part fite years' experience in engineering, operations and maintenance at nuclear or fossil-fuel power plants or similar facilities, with two yaars in a responsible supervisory position of such facilities and qualification as a licensed senior reactor operator.

  • Previously identified in CO Report No. 146/70-1.

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With the recent p'romotions of Messrs. Goodman and Potts, the test engineer positions as specified in Section 301, paragraph B.2. of the Saxton Final Safeguards Report and Sections N.1.a. (1) and (2) and Figures N.l.a.(1) and (2) of the Saxton Technical Specifications are presently vacant.

Mr. Swift stated recruiting efforts to fill these vacancies are presently in progress.

. l d.

A new position of maintenance foreman, who reports to the supervisor of reactor plant services and is equivalent in level to the radiation . protection engineer and the radiochemist was established at Saxton on , August 17, 1970 to relieve the supervisor reactor plant services of direct maintenance supervision duties.

Mr. S. Pekar, Senior Instrument Technician, is filling this position.

Mr. Pekar has 15 years of ' maintenance expa.rience with instruments which includes both hydraulic and pneumatic types.

He came to Saxton in 1962 as an instru=ent technician and since 1966 has been senior instrument technician, a position which has supervisory duties.

, ' e.

Mr. G. Reid, the site radiochemist, was granted a one-year draft j deferment and Mr. Swift does not anticipate Mr. Reid's being drafted in the future due to his age. A new radiochemist, Mr. R. Walton , ! joined Saxton in June of 1970 and was recruited as a replacement for Mr. Reid.

Mr. Walton recently graduated from Juniata College with a '

BS degree in chemistry.

f.

Since the last inspection five licensed reactor operators have departed the Saxton facility. Four operators ve.

to Three Mile Island and the other operator went to the General Electric Company. There are presently five licensed senior reactor operators and six licensed reactor operators on shift duty. The inspectors concluded that the licensee was in accordance with his Technical Specifications which require two " licensed reactor cperators at the facility at all tLaes.

The inspectors discussed the noncompliance aspects of the personnel changes enumerated in paragraphs b. and c. above at the exit interview.

Mr. Swift stated that their personnel changes were previously discussed with DRL in a May 1970 meeting and that DRL accepted the qualifications of Mr. Potts.

The inspectors wa.re informed that the licensee would submit a letter to DRL in accordance with paragraph 3.E.(2) of license No. DPR-4 reporting the recent personnel changes.* C.

Operations The inspectors reviewed the information in the operation log book, the On-Site Safety Committee minutes, and the SNEC Safety Committee minutes for April 20 - September 1, 1970.

In this review the inspectors noted that there was one planned shutdown and no unintentional scrams for this period.

Plant operations are su=marized as follows:

  • Subsequently sent to DRL on September 8, 1970.

774085 __

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May 14, 1970 Routins operation which had been maintained since the last inspeccion was curtailed when the plant was shut down for the scheduled fuel pin inspection by the Westinghouse Mobile Fuel Evaluation Team. At approximately 40 minutes after shutdown an unplanned release of radioactive gases occurred while venting gases from the primary coolant system to the gas compressor .j system. The licensee reported that an estLnated 7.32 curies of gaseous activity (predominantly Xe-133) were released to the environs over a four-hour period.* (See Sections P.3 and Q.4 of this report) 2.

May 27 - June 20, 1970 tj An IBEW strike against the Pennsylvania Electric Company occurred which involved the operators and technicians at the Saxton facility.

Mr. Swift ' stated that there were no significant effects on plant operation or maintenance as a result of the strike due to the fact that cold, shutdown, depressurized conditions existed in the plant, and the five removable subassemblies which were scheduled for inspection had been removed from the l core and placed in the spent fuel storage rack.

3.

June 26 - August 5,1970 . i Fuel pin inspection was conducted on the five removable subassemblies by the Westinghou=e Mobile Fuel Evaluation Team.

One or both of sister fuel pins from subassembly 504-4-33 were confirmed as leakers. (see Section G ' of this report.)

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4.

August 6,1970 Fuel subassembly 504-4-25 was dropped a distance of approximately 12 inches while being loaded into the spent fuel storage rack following its inspec-tion. Buckling damage to the subassembly can occurred which precluded its reloading into the core.

(See Section 0 of this report.)

5.

Aueust 19, 1970 Following completion of core loading and conoseal installation, primary system heatup was initiated utilizing pump heat and pressurizer heaters.

' 6.

August 20, 1970 , Separation of both 3/8 inch sensing lines to the pressurizer level indicator D/P cell occurred during primary system heatup at 2000 psig and 6360 F in the pressurizer. Excess flow check valves functioned as designed and there was no primary leakage.

(See Section E.1. of this report.)

, 7.

Au gus t 24, 1970 Af ter completion of the hot functional primary system pressure test at 2285 psig and 5120 F (50 psig above operating pressure), the reactor was made critical and flux mapping began.

  • Letter to DRL dated May 18, 1970.

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Aurast 26, 1970 An unplanned release of radioactive gases occurred while operating personnel were switching gas compressors to locate and isolate a mintee laak in the RWDF gas compressor system. The licensee reported that an estimated 0.034 curbs of gaseous activity (predominantly Xe-133 and Xe-135) were released to the environs over a five-minute period.* (See Sections P.4. and Q.4.

- of this report.

. 9.

Aurast 31, 1970 Load cycling operations were initiated and were in progress during the inspection. This operation consisted of cycling between 1007. and 40% of full power with 6 cycles being ~ performed en the day and swing shifts during a 5-day week. Mr. Swift stated that load cycling operations to test fuel performance and Westinghouse customer training would continue until the mid-life fuel inspection which is scheduled for November 1970.

' D.

Facilitv Procedures l.

The inspectors verified, in discussions with Mr. Swift that the licensee conducted an emergency preparedness plan drill on August 26, 1970.

. Mr. Swift stated that the results of the drill were satisfactory and that j the drill included phone checks to support agencies. This is a followup j itam from the last inspection.** ' 2.

The inspectors reviewed the calibration procedure for the Steam Generator Blowdown Monitor (RIC-5) and noted that the last calibration using known ,' source strengths (radioactive primary coolant water) was performed on octobe 14, 1969. The -inspectors also reviewed the calibration curve ' (uCi/cc vs CPS) for RIC-5.

The monitor is given a source check rcnthly utili 1ng a Co-60 source and a standard geometry set.up.

The calibration

point is 120 cps and the alarm point is set at 3 cps.

E.

Primary Systems 1.

Separation of Sensine Lines to Pressurizer Level Indicator D/P Cell In discussions with Messrs. Swift and Pekar it was learned that the pressurizer channel (LIC-2), which provides signals for continuous level indication, "On" "Off" high and low level alarms, low-level heater shut-off, and low-level scram, indicated an increase in level from 507 to 1007.

during primary system heatup (2000 psig and 6360 F in pressurizer).

The , plant was cooled down and investigation revealed that the Tylock fittings in both 3/8 inch stainless steel tubing lines connecting the pressurizer level column to a D/P cell had separated which caused the level indication on LIC-2 to increase to 1007.. There was no primary system leakage as a result of this occurrence because the excess flow check' valve (Chemiquip Type 50 FM 100) located upstream of the Tylock fitting in both lines, functioned as designed.

The separation occurred above the stainless steel

  • Letter to DRL dated August 31, 1970.
    • C0 Report No. 146/70-1.

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._ _ _ s. . . .. . O O ' ' - .- . ' - .. . " ' - _ 11 -- . salva block assembly, which is mounted in the D/P cell cabinct. The arrangenent is shown in Figure 1 attached.

Prior to system heatup the D/P cell had been removed for enlibration by uncoupling the D/P cell from the valve block assembly at the Swaglock fittings. Corrective action .caken by the licensee consisted of installing a new valve block assembly, . replacing the Tylock fittings with Swaglock fittings, recalibrating the j , D/P cell, and performing a hot functional primary system pressure (leak) test at 2285 psig and 5120 F.- _ 2; PrimarvCoolantkctivity

Examination of records for analyses of primary coolant disclosed that the highest coolant activity, as shown below, occurred on May 14, 1970, just prior to the scheduled shutdown for fuel inspection.

t ' Total Degassed Gaseous

t ! 180.03 uci/cc 16.80 uci/cc 163.23 uCi/cc ! Since returning to operation on August 24, 1970, the total coolant activity has decreased to about 14 uCi/cc. The reduction in coolant activity resulted from the removal of failed fuel pins from the core.

I F.

Reactivity Control and Core Pnysics ' ~ Core Power D!stribution ~ ' 1.

The in-core monitoring program was discussed with Mr. Goodman, and based ' on weekly and monthly Westinghouse reports, Saxton projects ahead their i operating power limits (currently 22.80 Mwt) based on peak pellet burnup ' (MWD /MIM) and peak pellet. power (Kw/ft). The inspectors reviewed operating data taken on August 31, 1970 at an equilibrium power of 22.77 - Mwt. The data which is su=marized below indicates the licensee is operating within the established Technical Specification limits.* Peak Assembly Core Position Peak Pellet Power Pellet Burnuo 97,233 Pu. Loose Lattice A9 in D-4 21.57 Kw/ft --- 25, 2 8 3. 7 5 -- - 9,03:.21-722 ', U Load Follow A7 in E-3 18.34 Kw/ft 2.

Control and Safety System ' The maximwn time from scram initiation to scram completion was reviewed by the inspectors. Maximum tbne allowed by the Saxton Technical Specifica-tions is 1.5 seconds. This measurement is the total of the control rod drive scram speed and scram circuit response time. The last test of control rod drive scram speed was conducted March 23, 1970.

The slowest tLne recorded was 1.050 seconds for control rod No. 4.

Scram circuit response time was last tested July 27, 1970. The slowest response time

  • Section G.3.

. - --. W4088 .-

- r .: O O - - . . - < . . - 12 - - . \\ ' recorded was 0.225 seconds for the high main coolant tamparature scram.

The combined time, based on the slowest times, is 1.275 seconds for scram i initiation to scram completion, which is within the 1.5 seconds allowed by the Saxton Technical Specifications.* G.

Core and Internals _ . , During the last shutdown the Westinghouse Mobile Fuel Evaluation Team inspected the fuel pins in the five removable subassemblies.

The fuel pins were cleaned, the , pin diameters measured, and pictures taken with a TV camera.

Mr. Swift stated that - the general condition of the fuel was good with the exception of sister fuel pins . lj No. 504 and No. 505 in fuel subassembly 504-4-33, a peripheral assembly in core l position N-2.** These fuel pins were monitored together and one or both pins were

confirmed as leakers. Mr. Swift stated that one pin had blisters and the other had ]j a. hole approximately 1/4 inch in diameter on the cladding surface which tapered to approximately 1/8 inch in diameter at so=e depth into the pin. The exact location mi of the hole with respect to the experimental fluoride contaminated longitudinal i notches was not made available to the inspectors.

These fuel pins were subsequently shipped to the Westinghouse Post Irradiation Facility at Waltz Mill, Pennsylvania for an in-depth inspection. The reportability to the AEC (DRL) of the results of l these fuel pin inspections was discussed at the exit interview.

l I.

Auxiliarv Systems

1.

Piping Modification to Accu =ulator on Charming Pumo Stai%4 3e Header .. . The 1 inch, schedule 160, 304 stainless steel seamless pipe which connects the accumulator to the charging pump discharge header has been extended , approximately 12 inches by utilizing a 1 inch, schedule 160, 304 stainless , steel pipe coupling rated at 4000 psi and a 12 inch length of 1 inch, schedule 160, 304 stainless steel pipe. This modification was made to , permit easier access for maintenance. A review of the modification package revealed that the pipe and welding wire used met Code ANSI B31.1 require-ments, but the material certificate for the coupling was not available.

Mr. Swift stated that the certificate had been requested frem Westinghouse - but had not been received. The inspectors will pursue this matter during ' the next inspection.

The welding procedure, welder, and welding inspector, ' which were supplied by Westinghouse, were qualified to applicable code requirements. The velds were L.P. tested in accordance with applicable code requirements and the system was hydrostatically tested to 3750 psig in accordance with the Saxton Final Safeguards Report.*** The design system pressure is 2500 psig.

  • Section N.4.f. (4).
    • Fuel pin description given in:

Safeguards Report for Saxton Core III, Table 2.1-3; Appendix C - Change Report 18; and in Saxton Technical Specifica-tions, Paragraph F.3.h.(2).

      • Section 206.E.

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Containment Bridge 20 Ton Crane Mr. Swift informed the inspectors that on July 1, 1970, a new cable (9/16 inch diameter Roebling Wire Rope) was installed on the bridge crane, the Ibnit switches set, and the gear box, gear box to drive motor coupling, drum to gear box coupling, and drum coupling inspected by the factory technical representative. Due to an operator error, the old , crane cable was sheared when it was inadvertently allowed to slip off ,the drum and wrap around the drum axle. The On-Site Safety Committee , reviewed this occurrence and reconnended that during use the crane be . attended by at least two SNEC personnel, one operator and one director.

. Mr. Swift stated that Westinghouse's chief crane inspector load tested i the 20 ton bridge crane to 111% of design load on July 6,1970. The ! inspectors reviewed both the inspection and load test reports and noted that the reports appeared to be complete and in order. The inspectors i also noted that the crane inspector rececmended replacement of the four lifting slings.

Mr. Swift stated that the slings had been ordered.

Testing of the 20 ton bridge is a followup item from the last two inspections.* _ ' ' O.

Fuel Handling l

Licensee correspondence with DRL** and subsequent discussions with Messrs.

Swift, Goodman, and Potts revealed that after inspection by the Westinghouse Mobile Fuel Evaluation Team, fuel subassembly 504-4-25 was dropped approximately 12 inches while being loaded into the spent fuel storage rack. Examination by the licensee , . revealed that the fuel pins were not damaged, but the 0.019 inch thick perforated l stainless steel can showed buckling damage about 4 - 5 inches frem the top and 3 - 4 inches from the bottom of the can.

It was decided by th'e' licensee that this subassembly would not be reloaded into the core.

(See previously referenced letter).

  • The subassembly fell when.the support tube separated from the protecrive sleeve during normal handling.

It was determined by the inspectors that this occurrence could.have hoppened at any time during the handling period due to an apparent failure in the locking device and that this irradiated subassembly could have been dropped a distance of up to 14 feet.

An illustration of the mechanical fitup for lifting a fuel subassembly is givdn in Figure 2.

The locking action occurs when the locking screws are turned 90 degrees by an operator after the protective sleeve is lowered into position over the guide tube. The fitup occurs prior to raising the water level above the top of the reactor pressure vessel head in preparation for fuel handling.

Mr. Swift stated that a large force is required to turn the locking screws but in this case the operator noted that less force than normal was required during the fitup.

Mr. Swift indicated that both screws apparently rotated during subsequent fuel handling, and that the fuel handling was considerably more than normal due to the inspection of the fuel pins. A review of the draft On-Site Safety Cecmittee minutes for the August 20, 1970 meeting indicated that this occurrence was not reviewed by the committee. This matter was discussed at the exit interview.

774090

  • C0 Report Nos. 146/69-4 and 146/70-1.
    • Letter to DRL, dated August 18, 1970, Technical Specification Change Request #37.

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Radiation Protection 1.

Monitoring Equipment The inspectors reviewed Saxton's surveillance testing records and i - verified that the radiation monitor circuits had been tested on monthly intervals for the peried January - August 1970 and that the se:rg-annual i 24-hour level test to verify the inner tank integrity of the liquid waste storage tanks had been performed and satisfactory results obtained as required by the Saxton Technical Specifications.* , 2.

Personnel Monitoring ' Examination of personnel monitoring records for the period of April - August 1970 disclosed that the highest individual exposure was 1350 mrem

in are month and 1785 mrem in one quarter. The inspectors verified that the licursee had determined individual accu =ulated occupational doses to - the whole body as required by 10 CFR Part 20.101 and 20.102.

3.

Personnel Monitoring for May 14, 1970 - Un planned Release (_ Radioactivity ** The licensee reported that the maximum concentration of gaseous radio-activity to which personnel were exposed while in the gas compressor room was 3.3 x 10-4 uCi/cc (predeminantly Xe-133) and that based on consultation with their medical-radiation consultant and as a precautionary measure, ' bionssays on all personnel involved were obtained.*** The events of this release were discussed with Messrs. Swift and Beale.

It was learned that the highest external exposure recei 'l during the release was 10 mrem as ' ' ' - indicated by one individual's pocket dost=eter.

Bioassays (urine and fecal)- - for the personnel involved were reviewed by Saxton's medical-radiation consultant, the On-Site Safety Co=mittee, and the'SNEC safety committee, and it was determined that there was no evidence of body burden.

A review by CO:I of the bioassays did not indicate any significant body uptakes.

4.

Personnel Monitoring for August 26, 1970 Unplanned Release of Radioactivitv**** The licensee reported a small release of gaseous radioacchdty (0.034 Ci of predominantly Xe-133 and Xe-135 over a five minute period) and that there were no personnel exposures.***** In discussions concerning the events of this release with Messrs. Swift and Beale it was learned that radiation readings, smear surveys, and gas samples taken in the gas compressor room were normal. Nasal swipes and smears were taken on personnel in the area and the results were negative. As a precautionary measure, the licensee had bioassays (urine) performed on personnel in the area. The results of the bioassays had not been received from Tracerlab during this inspection and will be reviewed by the inspector during the next inspection.

(O b [ NY' M M 2 q r, orandum No. 146/70-B.

      • Letter to DRL dated May 18, 1970.
        • Inquiry Memorandum No.14ti/70-C.
          • Letter to DRL dated August 31, 1970.

774031 __... - - ' O O ' - . .. . . . - -. . - 15 - '

. . Q. Radioactive Waste Systems . > 1.

Lfquid Waste Liquid release records for the period January - Au;ust 1970 were revietred by the inspectors and no indication was found cha; releases have exceeded l applicable limits.

Following is a su= mary of liquid waste releases for this period.

i , I Beta-Ga=ma Tritium , 0.007521 Curias 6.7224 Curies , I 2.

Gaseous Waste Gaseous release records for the period January - August 1970 were reviewed , ,

by the inspectors and a su= mary of gaseous waste releases for -this period and the tLee averaged percent of Itnit are given below: Xe-133 & Xe-135 % of Limit I-131 % of Limit . 1708;2 Curies 6 87.

0.011071 Curies 0.17% Release,1Lnits of 10 curies / year' of I-131 and 3750 curies / year of krypton . and xenon are stipulated in the Saxton Technical Specifications. Releases for the period of June - August 1970 totaled about 6 curies and with the removal of the failed fuel pins it would appear that the applicable release , limits for calendar year 1970 will not be exceeded.

,.. '% 3.

calculation of Gaseous Activity Concentration The licensee reported by telephone to CO:I on June 1, 1970 that Westinghouse had discovered errors in the ge=ma abundance factors being used by Saxton personnel to calculate gaseous activity concentration.

In view of this finding Westinghouse reviewed all of Saxton's counting procedureu to verify their adequacy for determining activity releases.

The licensee stated that gaseous releases for the period January - May 1970 would be corrected and . published in the Saxton Monthly Report for May 1970.

In discussions with Messrs. Swift and Reid'the inspectors noted that the following corrections had been made to the ge=ma abundance factors used in the Noble gas counting procedure.

Isotope ,Y - Abundance Factor Old New Xe-133 1.0 .37 Xe-135 1.0 .91 Kr-87 1.0 .85 Kr-88 1.0 .35 77JiOS2 -... - - ... ._. -

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' - , .- - 16 - . , After reviewing the gar:ous activity release records and the Saxton Monthly Report for May 1970, the inspectors concluded that the licensee had . corrected gaseous releases for the period of January - May 1970, and . ,that after May 31, 1970, all gaseous releases were calculated using the .new gamma abundance factors.

, The inspectors also noted that past liquid releases had not been

corrected by the licensee and that no records were available to substantiate the basis for not correcting these releases. The inspectors questioned the lack of definitive basis on the part of the licensee for not correcting the liquid releases at the exit interview. Mr. Swift stated that an j analysis would be performed to ' substantiate the basis for not correcting past liquid waste releases.

,j 6.

tinnlanned Gaseous Activity Releases to the Environs l Msv 14, 1970 Release * ! The inspectors reviewed the events of this release with Messrs. Swift, ' Beale and Pekar. It was noted that during the release the stack monitor

(RIC-3) was pegged full acale (reading > 1000 cps) for a period of 3 - 4 . hours. Examination of the calibration curve for RIC-3 indicated that a reading of greater than 1000 cps corresponds to a gaseous activi v j concentration of greater than 1 x 10-3 uCi/cc.

Based on these findings it appeared to the inspectors that the Saxton Technical Specification j release limit for an instantaneous concentration of 1 x 10-3 uCi/ce, , i excluding I-131, when averaged over a 15-minute period ** was exceeded during this release. This matter was discussed at the exit interview.

The licensee informed the inspectors that further investigation of ! No.1 gas compressor vacuum regulator valve (Fisher Governor Type Y600)

disclosed that, contrary to the information given in their report to i DRL, the diaphragm had not ruptured but rather the adjusting rod for setting desired system vacuum was.found adjusted in such a manner that ' the center hole in the diaphragm had lost its seal and thus provided the leakage path.

'The inspectors also verified that the release of an estimated 7.32 curies of gaseous activity which was predominantly Xe-133, as reported by the licensee, was calculated using the old ga=ma abundance factor for Xe-133 of 1.0, and based on the new ga=ma abundance factor for Xe-133 of 0.37 , would yield a release of an estimated 19.76 curies.

- The release was reviewed by the On-Site Safety Committee and the SNEC Safety Committee and the following precautionary actions have or will be taken: -

  • Inquiry Memorandum No. 146/70-B.
    • Section I.8.d.

774093 .. .

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The surge tank venting procedure was reviewed, however no changes , were made because the release was caused by mechanical failure.

. b.

Routins pressure testing of vacuum regulating valves and the gas compressor system.

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Review proposed isolation procedures.

! i d.

Instruction of SNEC personnel in the proper use of a Scott Air Pack.

j ' August 26, 1970 Release * t The events of this release were discussed with Messrs. Swift, Goodman, Pekar and Beale. It was noted that for approximately five minutes during this release RIC-3 recorded a reading in excess of 1000 cps or a concentra-tion greater than 1 x 10-3 uCi/cc. Based on the total activity released, time duration of the release, and dilution factor in the stack, the , inspectors concluded that the Technical Specification instantaneous

I release Itnit of 1 x 10-3 uci/ce, excluding I-131, when averaged over a 15-minute period probably was not exceeded.

The inspectors concluded that the release occurred as reported by the licensee, but noted that leakage back through gas compressor systems involved passage through six check valves prior to release. The RWDF (both gaseous and liquid systems), which is composed of carbon steel valves and piping, is subject to corrosive attack from liquid waste, i i especially boric acid. The inspectors reviewed maintenance records ~ which indicated that six check valves and six globe valves in the gas ' compressor system had been inspected and repaired where possible with deficiencies noted. The licensee indicated that replacement parts for - defective valves had been ordered'. The general condition of the RWDF was discussed at the exit interview.

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  • Inquiry Memorandum No. 146/70-C.

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. FITUP FOR LIFTING FUEL SUBASSEMBLY Figure 2 - ... .. - _.. -. }}

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