NUREG-0837, Discusses Data Suggesting That Factors Other than Operation of & Releases from Facility Affecting Results of Environ Monitoring Program Re Resin Release in June 1982.Higher Dose Readings & 1982 Release Coincidental.Related Info Encl
| ML20154G588 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 09/25/1986 |
| From: | Fuhrmeister R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Kane W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20154G583 | List: |
| References | |
| RTR-NUREG-0837, RTR-NUREG-837 NUDOCS 8805240433 | |
| Download: ML20154G588 (6) | |
Text
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MAR.22 '86 11:16 e n era n
. )
\' . .
l SEP 2 5 396 i
William F, Kane, Director, Division of Reactor Projects j MEMORANDUM FOR:
THROUGH:
Harry B. Kister, Chief, Reactor Projects Branch No.1. ORP 4 Jack Strosnider, Chief, Reactor Projects Section No. IB, ORP FROM:
1Roy L. Fuhrmeister, Reactor Engineer, RPB Ho.1, DRP l
]
SUBJECT:
PILGRIM RESIN RELEASE IN JUNE 1982 During recent public meetings in the vicinity of P1ynouth, Massachusetts there have been numerous references to the resin release at Pilgrim in June of 1982.
These references have most often been made by Mr. Abbott of the Plymouth County Nuclear Information Group in the manner of "the accident in 1982". A great deal has been made of the increased dose measured on a particular Thermo-Lumi-nescent Dosieeter (TLD) during the sum-er and autumn of that year. In order to determine if there was any credence to the claims that the Pilgrim resin release contaminated the environment as far away as New hampshire, a number of TLD data points were extracted from the NUREG 0837 series and plotted on a common time line. An explanation of the data points selected from random plants, a tabula-tion of the data, and a plot on the common time axis are attached.
It is interesting to note that during the first half of 1984, while the plant Also vas shut down, TLD 1 from Pilgrim showed striking increases in the dose.
of note is the fact the TLD 49 from Pilgrim, located in Weynouth, Massachusetts, shows a consistently higher dose than TLD 13, which is caly 0.7 miles f rom the plant.
and quarterly release data As areaalso check on TLDin1, included the the plant operation time-lineNo correlation with plant activities is f!gure.
apparent. In fact, the high reading in early 1984, with the plant shut down and no releases being made is inconsistent with Mr. Abbott's contentions.
In conclusion, it can be seen that the off-site dose in the vicinity of Ptigrim Nuclear Power Station followed the general trend of the other sites in the Northeastern United States. This trend includes a significant drop in doses during the first quarter of 1982. This drop, if narr oly construed, could lead l
one to the conclusion that the second quarter 1982 dose was significantly higher.
This would be an erroneous conclusion, since second quarter 1982 dose is lower than the fourth quarter 1981 dose. In general, it appears that from mid-1981 to ;
1 eid-1983, Eastern Massachusetts dose data followed the decreasing trend evidenced In fact the doses in Eastern Massachusetts, across the Northeast United States.
including those measured around the Pilgrim site (with the exception of TLD No. I which is exposed to turbine "shine"), were on the order of 70% of expected natural background throughout the period. First quarter 1983 doses show a dramatic drop in Eastern Massachusetts, despite a major release from Pilgrim during that time It is also wortby of note period (13,200 ci, higher than the 1982 release).
that with the exception of the third quarter 1981, Weymouth, Massachusetts doses j2 33 080429 U CK obooon93 PDR
s an. , c. c. . . . . . - . . .....
't . .
2 were higher than those recorded only 0.7 miles from the site. This suggests ,
factors other than operation of and releases from Pilgrir are affecting the results of the environmental monitoring program. This also shows that the 1982 resin release and higher dose readings are strictly coincidental.
The dose readings on TLD 1 are in the range of 1 to 3 times the expected back-ground levels for the area. The cause of the elevated readings was originally thought to be "turbine shine". The 1984 data do not suoport that cenclusion, and further information on plant activities in 1984 is being developed. Parti-cular interest is being paid to temporary on-site storage of materials removed during the recirculation piping changeout.
Attachment 6 shows typical expected doses.
?"e^
Roy L. Fuhrmeister Reactor Engineer, RPS IB DRP Attachme nt s :
- 1. Explanation of Data Points
- 2. Tabulation of Data
- 3. Plet of TLD Exposure Data 1981-1984
- 4. PNO-!-82-42
- 5. PHO-I-82-42a
- 6. Extract from Health Physics and Radiological Health Handbook cc:
H. Kister J. 5trosnider r L. Doerflein K McBride i
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P24 MAR 22 '88 lit 17 PILGRIM NRC RESIDENT Ro r c7 7;.: F Pwset Dern Fa h81-1iO
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P25 l MAR 22 '88 11:18 PILGRIM HRC RESIDENT i,si.:*:. E .. " .: : :.C . . .; . : '. ' . : . ' . : . ' - u s ,,, ,., s , e:. a . sve- :- st.ta:es a. - cot.e s se ,ss opt :arce ine tn'e r e. e s .: c. . . . s.., c.sion or .. 1.asion anc o eas cait a n v.4 : : . . . . :., . PtigrlehuelearPower$tation Pty outn. Massacnusetts u c e n s u a r p n , :. ,, v o .. . . p geco p ,; OR 50 N3 heu ficar e. < %;. s : .
~ Alert l Site Area Errager: 4 1
-** 4 General bergency I
- hot AppItCaDlb
. Itt.Dt$t 0F SPDtf K51M f
- approainstely 1300 on June 11, 1982 spent msin was found on th cour.d te.i e .'
turMee ts11 ding. 5ebseouset surveys identified contamination of tne roofs r
- te. i hrthms. Reector, Off-f,as and Ar Tube Buildings. Contamination ws- also four grused wittir. the sita controlled amas. Contamination levtl. ranged frn 0 gunfM0 cud witA maximse contarination of up to 100.000 dom / W. u Ca i .
, i.
ans M is of the resta teuntified primertly long lived radionuclides (Lc-(,.
. Cs-EM and lei-54). ,
l All persnne..' ,a rs tw - to sentas6eetten uns identified off.stte or in storn drains. - frtshed prior to emiting the sita and no personnel contanination has beer i-he resta asy have been released through the reactor building vent duct 'i - nt . The licensee has foune
" ~~ ' to te atmosphere og en elevetten of approxlestely 100 f t. -
apprestestely 10 f t of resin in the Staney Gas Irvatrent Syster inlet elenu . (k ,r. eeurce of the resta is being investigated. Three radiatioet specialisu have bece dispeteed to the site ta evaluate the r6diological aspects of the occurreno . hY ... Stedla laterest is spected dw to pelic interest in the facility. T h.- liter .- i-eensleering issateg a press release. The h8K does not plan to issue a press r h .. .
- but Will respond to media inquiries. The Commonwealth of Massachusetts t..is ba r
,M,. teferued. ,
his PR is emneet as of 4:45 P.M., June 11, 1982.
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nnn a c.c. s s e s c r s 2.~ w . n r. .. - n c. ; . w . DC5 Nr '0293-820611 /' -[ j Date: . ne 14, 1982 PRELININARY NOTIFICATION OF EVENT OR UNUSUAL OCCURRENCE-PNO-I-82 42A
- This preliminary nottfication constitutes EARLY notice of events of P0$518L1 safety or public interest significance. The information is as initially received without verifi-cation or evaluation, and is basically pil that is known by the Region I staff on this date.
Pilgrim Nuclear Power Station Facility: P1, mouth, Massachusetts Licensee Emergency Classification: i DN 50 293 Notification of Unusual Event Alert
$tte Area Emergency General Emergency x Not Applicable
Subject:
RELEASEOFSPENTRES!N(UPDATEPNO!-82-42) Surveys of the entire site within the erotected area and surveys of selected areas of the licensee controlled area were made within 3 hours of the identification of the spent resin release. The licensee's onsite surveys identified two contaminated pavement areas which were barricaded and posted. Surveys confirmed contamination of the Turbine, Administration Augmented off-Gas and Re-Tube Building roofs. The Reactor Building Roof was found to be free of contamination. The licensee's offsite survey included surveys of cars, parking lots, shorefront, and security access areas. No contamination was identified. Routine enviromental air samples covering the period June 1-15, 1982 were counted. Nothing unusual was identified. Because of the size and weight of the resins, no offsite airborne i release of the beads appears to have occurred. This was confimed by air samples collected during clean up of the contaninated pavement areas which when counted indicated background and the identification of resins only on roof-tops under the Reactor Buiding Vent. Prelimi ary samples of stem drain residue have been counted with no contamination identified. All contaminated ventilation ducts have been vacuumed clean. A duct surveillance program has been established to identify any additional resin accumulation. s The licensee believes the resin entered the ventilation ducts from the condensate dominer-aliter system during resin backwashing via the Cation Regeneration Tank Vent. In addition, resin from defective condensate domineralizer vent valves may have also been released prior to their repair during the September 1981 -March 1982 refueling outage. The resin appears to have been released from the Reactor Building Yentilation Exhaust System which vents above the reactor building roof, prior to the repair of defective filters in this system in September 1981. . I I The licenses has suspended all transfer operations which could result in further resin releases to ventilation ducts and has initieted additional environmental sampling. The
- licensee's actions were monitored by three Rqion I Radiation Specialists throughout the
! teekend. Region I will issue a Confirmatory ution Letter to address olanned licensee
- corrective actions. The itcensee is continuing to review the source and cause to detemine what pomanent corrective action will be needed. The Resident Inspectors are closely 1
following licensee actions concerning this event. l Media interest has occurred. The licensee has responded to media inouiries but does not i plan to issue a press release. The NRC will respond to media inouiries but does not plan to issue a press release. This PN is current as of, 11:00 aa. , , June 14, 1982. ,
- /
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P28 MAR 22 '88 lit 20 PILGRIM NRC RESIDENT l . [ hh l Table 1.5. Sune.ary of average annual per capita dunes to whole U.S. population ' 9 _-_ I Ave. per capita dose (area / year) Source Natural background 31 Cosmic 68 Terrestrial 4 Tech. inhanced 103 Sub-total Man-made Medical ' 77 K-ray 14 Nuc. Med. 91 Sub-total 4-5 Nuclest weapons Nuclear power
<1 consumer products 0.5-1.5 8
Sub-totsi
-200 i) Total Table 1.6. U.S. general population collective dose estimatea - 1978 (From liologic Ef f ects of loniting Radiation. Report of the Science Work Croup of the Interagency Task Force on Radiation, Department of Realth, Education and Welf ara, June,1979)
Person-reme per year Source (in thoveande) Natural background 20,000 1,000 Technologically enhanced 16,000 Coaling arts Nuclear weapons 1,000-1,600 , Tallout ) Weapone development, testing 0.165 and production 56 l Nuclear pewsr ' 6 Consumer products
3;,p 22 ' 88 11: 20 FIL6 RIM f 4AC st OWE'4I Ei# _
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Table 1.7. Annual per capita does from naturel radioettivity , l
, Source Variab111ty Dose (area / year)
Cosmic Average
- 31 Rock sountain states 60 *80 Jet flight - trans continental 2.5/ Trip Terrestrial (external) Average
- 40 i Colorado 75-140 l (internal) Average * (sonada) 28 !
I Lung 100-450 Tech. inhanced Average
- 4 Total g l l
' Average whole-body does to the whole population.
Uncorrected fer shielding of structures (reduce cosmic by 101 and terrestrial by 201). Self-shielding by body further I reduces dose. l Table 1.8. Radiation desse frem medicel radiation" Source Mean active bone-marrow dose , l Ave. per capits dose tres/exas (ares / year) Diagnostie x-rays chest :-ray 10 Upper CI $00 tower C1 900 skull 80 Full mouth (dental) 9 8 Sub-total 77 Radio pharmaceuticale Dess (aras) to organ Specified/exas II33 (function) Thyroid $000 Whole body 30 993 Tc Vhole body 180 i II Ke Whole body 5
. Whole-body equivalent to (
whole population 14 Total g "Doesn't include therapy
,
- Based on whole population (suposed and uneaposed) g
'G5D is 20 ares / year (CSD is the Genetically Scientific Dose)
. ThQUESTf0'7 NUREG-0837 Vol. 2, No. 3 1
NRC TLD Direct Radiation Monitoring Network 4 Progress Report
- July-September 1982 4
U.S. Nuclear Regulatory Commission NRC Region i ) - 1 j F. Costello, T. Thompson, L Cohen , p>* * *eu, e g o...* j 4 l
. - - - - . . . .-_ . - _ . . . . . _ . _ _ . . . . _ . _ . _ - . . . . _ _ _ _ . . . _ _ _ . _ _ _ _ . . . _ . _ _ _ . . . _ _ _ _ _ _ _ . _ . ~ _ . , . _ _ _. . . _ - - ~ . _ . - , - -
-- - ~ _ _ . . .
l P!LGRIM JppjREg7 rag!AJgQN ENV L RgNMl:HJR MgNITORING FELDk!Mk bleIbhbblbObbfDMhk GROSS EXPOSURE RATE HRC LOCATION EXPOSURE (mR) mR/Std.0tr. STAT 10H AZ1MUTH/DIST (mt.) +- Std. Dev. +- Std. Dev. (deg.) 001 288 0.10 52.1 +- 1.3 46.4 +- 1.1 0.20 19.6 +- 1.1 17.5 +- .9 002 310 0.70 19.4 +- .2 17.3 +- .2 005 289 1.70 17.9 +- .9 15.9 +- .8 006 261 1 l 0.50 19.4 +- .0 17.3 +- .0 007 270 247 0.30 19.2 +- .3 17.1 +- .3 000 224 0.30 17.6 +- .3 15.7 +- .2 l 009 l 010 205 0.30 27.1 +- 1.2 24.1 +- 1.1 t i 22.2 +- .3 19.8 +- .2 011 184 0.03 l
.5 20.6 +- 4 i 012 159 0.40 23.2 +-
ll 0.70 15.0 +- .5 14.1 +- 4 013 146 155 1.00 17.4 +- .0 15.5 +- .0 014 18.1 +- .2 016 136 1.30 20.3 +- .2 212 0.60 19.7 +- .0 17.6 +- .0 016
.3 13.0 +- .3 019 232 1.00 14.6 +-
1.60 16.8 +- .1 15.0 +- .1 021 256 13.7 +- .5 022 130 2.50 15.4 +- .5 146 3.40 15.3 +- .8 13.6 +- .7 023
.3 13.9 +- .2 025 148 1.50 15.6 +-
.2 13.3 +- .2 026 180 1.30 14.9 +- l
i-P!LGRIM 8 7 reg QH EN HMENTA N!TORING g rag!A !yLRI;bbhfDh hEL k!M h Ih l HRC LOCATION GRO$$ EXPOSURE RATE STAT!0H A21MUTN/D!$T EXPOSURE (mR) mR/Std.Qtr. (deg.) (mt.) +- Std. Dev. +- $td. Dev. 427 231 1.89 16.3 +- .7 14.5 +- .7 938 153 2.20 17.1 +- .1 15.2 +- .1 - 031 179 2.50 15.2 +- .8 13.5 +- .0 03? 217 2.60 13.9 +- .4 12.4 +- 4 433 234 2.50 16.0 +- .2 14.2 +- .1 a 837 264 4.20 17.9 +- .1 15.9 *- .1 i 839 155 5.30 13.3 +- .2 11.9 +- .2 040 272 4.66 16.2 +- .1 14.5 +- .I 843 291 5.60 18.2 +- .6 16.2 +- .5
; 045 - - 13.9 +- .8 12.3 +- .0 i 47 3.i 26.2 i... .- .. 16.. .- ..
q 048 301 26.2 17.7 +- .2 15.0 +- .1 i 849 301 26.2 17.2 +- .1 15.3 +- .1 l l CCHMENTS: STATION 1 IS ON L3CCNSEE PRDPCRTY (P!LCRIM OVERLOOK ARCA). AtttBB 15 CONTROLLED i at } O Ii .. i l s l 4
.m 6 kbkkkEPERIOD$20630-521008 101 DRYS TLD D!PECT RALIATION ENVIRONMENTr.L MON!TORINO 9VFR. EXPOSUPE += Std.Dev. 4 IN GROUP RZIMUTH (deg.) (mR/Etd.Ctr.) 0 No DRTA+-NO DAT R 346.73-11.25 (N) 0 11.25-33.75 (NNE) NO DRTA+-No DATR NO DRTA+-No DATR 0 33.75-56 23 (NC) NO DRTA+-N0 DATR 0 36.25-78.73 (ENE) NO DRTA*-NQ DATR 0 76.75-101.25 (E) NO DRTA+-NO DATR 0 101.25-123.?5(f5E) 4 123,75-146.25fSE) 14.9 +- 2.2 14 6 . 25 -168. 7 5 ( 55 C ) 15.4 +- 3.3 5
- 15. 5 +- 3.7 3 169.75-191.25(5) 20.9 4- 4.7 2 191. 23 -213. 7 5 ( 55 H )
1.3 5 213.75-236,25(SW) 14.0 4-16.0 +- 1.5 2 236 25-258.75(W5W) 4 256.75-281.25(W) 15. 5 +- 1. 2 26.7 t- 17.l 3 28 L . 25-303.75 ( WNW) 303 . 75-22 6. 25 ( NH ) 17.5 t- D.0 1 ; i NO DRTAe-N0 DRTR 0 326.25-348.75(NNH) - ) AVCR . CYP 05URC +- St d . De v. 4 IN GROUA DISTRN;E(mi) FROM THE REACTOR inR/5td.0tr.) 18.3 *- 7.4 19 9-2 8 2-5 14.1 +- 1.1
>5 14.1 *- 3.1 2 UPWIND CON TROL D AT A 15.? * .3 3
.?
{ NUREG-0837
". Vol. 2, No.1 NRC TLD Direct Radiation ~
l Monitoring Network Progress Report January - Vlarch 1982 U.S. Nuclear Regulatory Commission F. Costello, T. Thompson, L. Cohen paase.q n l
\ q l %
'M kh: E FEP!0D ?11222-?20415 115 DAYS TLD DIF.ECT F ADI ATIOrt ENV I F 0tiMEtiT AL f10111 T O P I t1G
).
FCIt1U TH i. d e g . i AVER. E 8 P03UF E +- iit d . De v .
- If1 GPOUF (mP/Std.Qtc.)
348.75-11.25 t ri a O.0 +- 0.0 0
~
').25-33,75 ( fille ) 0.0 +- 0.0 0 f
75-56.25 ( tie ; O.0 +- 0.D 0
.25-79.75 ( E f1E ) O.0 +- 0.D 0 78 75-10l.25 LE4 0.0 +- 0.0 0 101.25-123.75tESE 0.0 +- 0.0 0 i
123.75-146.25 3E:' 15.3 +- 2.3 4 146.25-159.75i55EJ 16.3 +- 2.9 6 168.75-191.25t5) 14.9 4- 1.3 3 19 t . 25-213.75 t S5H i 18.8 + .9 2 213.75-236.25isWe 13.B 4- .? 5 236.25-258.75(64564) 15.9 + .6 2 258.75-281.25(612 15.8 +- 1.3 5 28 l . 25-303. 75 (64td64 5 18.D +- 5.9 6 303 75-326.25iNWJ ]$.8 +- 0.0 1 3 2 6. 2 5-3 4 8. 7 5 ( titiH ) 0.0 +- 0.0 0 DI ST A ric.E ( m i > FP0ti THE REACTOR AVEP.EXP05UFC t- Std.Det + Iid GPOUP fnR/Std.0tv.1 0-2 16.9 +- 3.7 19 2-5 15.0 +- 1.6 10
>5 14.7 +- 1.4 5 l J
1* l j PI'LGRIN T Ii DIFEc7 FADIATIQH EHVIF0HMENTA MQHIT0FING 14*5- iiv415 115 iA h F F THE #EPIOD it,0105
, F ELD f!ME ** :: 0 4 u e. vc D Yi i
HFC LOCATIGH INTEGRATED EXPO!UFE FF STAT!0rt A: IMUTH'D!iT EXPOSUFE(niR) n. F St d.0t r . (deg.) ' ni s . )
, +- Std. Dev. +- St d. Dev.
027 231 1.90 18.2 +- .3 14.3 +- .3 030 153 2.20 20.5 +- .5 16.1 +- 4 031 179 2.50 17.8 +- .1 13.9 +- .1 032 217 2.iO 16.8 +- .1 13.1 +- .1 033 234 2.50 17.8 +- .3 14.0 +- .2 037 264 4.20 20.1 +- .0 15.7 +- .0
)
0?8 152 3.50 24.1 +- .3 18.9 +- .2 039 155 5.30 16.3 +- .1 12.8 +- .1 040 272 4.60 18.5 +- .6 14.5 +- .5 i 042 281 4.60 18.8 +- .6 14.7 +- .5
!! 043 291 5.10 21.0 +- .1 16.4 +- .1 E!
045 - - 15.9 +- 4 12.4 +- .3 hc 047 301 26.2 17.8 +- .0 13.9 +- .0 { 048 301 26.2 19.9 +- 4 15.6 +- .3 kl 049 301 26.2 19.0 +- .7 14.9 +- .6 1 0 050 CTL TLD 17.1 +- .0 13.3 +- .0 c i 1 I l 1 N COMM ENTE: ETATIGH t IE ON LICENSEE PROFERTY (PILGRIM OVEPLOOK AFEA). ACCE55 15 C0flTPOLLED I 1 l l 4 l l
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t
- PILCRIN T EARIAT!?H EH MnHITOP1HG F D
& DIPECT T H E P E F I O u y l l *l 0 5 - 5,,U F ELD TIME e .' u .040e D YS41 % v} 4} 522-Ayl50HMENTA AYi HPC LOCATION INTEGRATED EXPOSUFE F**~
STATION A~IMUTH/DIST EXF05URE(mF) mF /St d.0t r -- (deg.) ( ru s . :. +- Std. Dev. +- Std. De 001 288 0.10 38.2 +- .8 29.9 +- .6 002 310 0.20 21.5 +- .7 16.8 +- .6 _ 005 289 0.70 21.9 +- .3 17.2 +- .2 - 006 261 1.70 20.6 +- .1 16.1 +- .1 - 007 270 0.50 22.7 +- 4 17.7 +- .3 008 247 0.30 20.9 +- .2 16.4 +- .1 009 224 0.30 18.9 +- .1 14.8 +- .1 010 205 0.30 24.8 +- .0 19.4 +- .0 ~ 011 184 0.03 21.0 +- .1 16.4 +- .1
~
012 159 0.40 26.1 +- .4 20.4 +- .3 013 146 0.70 17.3 +- .2 13.6 +- .2
~
014 155 1.00 19.8 +- .5 15.5 +- 4 016 136 1.30 23.8 +- .0 18.6 +- .0 018 212 0.60 23.2 +- 1.0 18.1 +- .8 019 232 1.00 16.6 +- .3 13.0 +- .3 021 256 1.60 19.8 +- .1 15.5 +- .1 022 130 2.50 19.1 +- .0 14.9 +- .0 023 146 3.40 17.8 +- .6 13.9 +- .5 025 168 1.50 17.8 +- .6 13.9 +- 4 026 180 1.30 18.3 +- .3 14.3 +- .3 o
,0 __
i NUREG-0837 Vol. 2, No. 2 NRC TLD Direct Radiation Monitoring Network Progress Report Apnl-June 1982 U.S. Nuclear Regulatory Commission NRC Region i F. Costello, T. Thompson, L. Cohen v'*"'%, j s ft
q r-- - - - - - - - - - hh PERIOD 800?25-820712 110 DAYS I TLD DIFECT RADIAT10ti EtiVIFONMENT AL MotlITOPIrlG AVEP. E s F 05 U RE + - S t d . 0+ v .
- Iti GPOUP RZIt tU TH (dag.)
(mR/Std.Qtr.) p h i 0.0 +- 0.0 0 0 348.75-11.25 ( ti) 0 i 11,25-33.75 itetJE) 0.0 +- O.0 0.0 +- 0.0 0 j 33.75-56.25 ( f1E ) 0 56.25-78.75 ( E tie i 0.0 +- 0.0 0.0 +- 0.0 0
# 78.75-10t.25 (E) 0.0 +- 0.0 0 101. 2 5 - 12 3.75 ( ESE )
1.1 4 123.75-146.25(SE) 16.0 4-17.9 +- 4.0 5 14 5. 25-168.75 t 55E ) 10.4 +- 5.0 3 168.75-191.25(5) 18.9 +- 2.9 2
~
191.25-2l3.75(55Hi 16.6 4- 2.2 5 213.75-236.25(560 17.3 +- 1.4 2 Y 2 3 6 . 25-25 8. 75 ( WS 61)
- 17. 9 +- 3. 4 5
. 250.75-201.25(W) 26.5 +- 13.8 3 ? 20 l . 25-303.75 (64tlH) 1 3 03 . 7 5-3 2 6. 25 ( tJH ) 18.0 +- 0.0 0.0 +- 0.0 0 3 26. 25-3 4 8.75 ( tit 4H )
AVCP.E~<POSURE +- Std.Dev. + Ild GPOUP 0 3 ET A ti'_E ( m t ) FPOH THE PEACTOR (nR/Std.0tr.) 19.5 +- 5.4 19 0-2 16.4 +- 1.6 9 2-5 16.3 +- 3.3 2 55 17.0 + .4 3 UPHItlD C0tiTP0L DRTA
FILGRIM P0HMENTA MgHITORING T g gEEgT P Agil ATgH ENVb04b!~hbo7bfhbD FIELDh1Mb LOCATIOH GROSS EXF05URE FATE HRC EXPOSURE (mR) mR/Std.0tr. STATION AZIMUTH /DIST (deg.) (mi.) +- Std. Dev. +- Std. Dev.
.5 14.9 +- 4 I
J 026 180 1.30 18.2 +-
'l 027 231 1.80 19.4 +- .3 15.9 +- .3 1
1 2.50 18.7 +- .2 15.3 +- .2
- 031 179 1
032 217 2.60 24.5 +- .1 20.0 +- .1 i .6 15.5 +- .5 033 234 2.50 18.9 +- j
.5 17.4 +- 4 037 264 4.20 21.3 +-
19.1 +- .5 15.6 +- .4 j 039 152 3.50 155 5.30 17.0 +- .1 13.9 +- .1 039 19.8 +- .3 16.2 +- .3 040 272 4.60 r 18.4 +- .4 15.1 +- .3 042 281 4.60 5.80 22.7 +- .2 18.6 +- .2 043 291 047 301 26.2 20.3 +- .1 16.6 +- .1 20.9 +- .5 17.1 +- .4 E48 301 26.2 21.3 +- 4 17.4 +- .3 049 301 26.2
- l. I .
l. i COMMENTS: e 5TATION 1 IS ON LICENSEE PRDPERTY (FILCRIH OVEPLOCK AFEA). RCCE55 15 CONTROLLED 1 I o 0 a . _ _ - _ ..
L i PILGRIN T pIREgT P Agli3T ijH EHg0NMENTAgMgNITORING FIELD 11Mk 51b4hhb107bhhbDA NRC LOCATION OROSS EXPOSURE RATE STATION AZIMUTH /DIST EXPOSURE (mR) mR/St d.0t r. (deg.) (mi.) +- Std. Dev. +- Std. Dev. 001 288 0.10 51.9 +- .5 42.5 +- .4 002 310 0.20 22.0 +- .1 18.0 +- .1 005 289 0.70 22.5 +- .6 18.4 +- .5 006 261 1.70 21.0 +- .1 17.1 +- .1 007 270 0.50 29.0 +- .3 23.8 +- .2 008 247 0.30 22.4 +- .0 18.3 +- .0 009 224 0.30 21.0 +- .6 17.2 +- .5 010 205 0.30 25.7 +- .4 21.0 +- .3 011 184 0.03 30.7 +- .9 25.1 +- .7 012 159 0.40 28.1 +- .2 23.0 +- .2 013 146 0.70 18.6 +- .3 15.2 +- .2 014 155 1.00 26.1 +- .0 21.4 +- .0 i l 016 136 1.30 20.0 +- 4 16.4 +- .3 018 212 0.80 20.6 +- .2 16.9 +- .1 i 019 232 1.00 17.5 +- 1.1 14.4 +- .9 1 021 256 1.60 19.9 +- .3 16.3 +- .2 022 130 2.50 18.4 +- .2 15.1 +- .2 1 023 146 3.40 21.2 +- .1 17.3 +- .1 025 168 1.50 19.1 +- .1 15.6 +- .0 f* i f,
UtiiT ED S T ATEs
# '% ENCLOSURE 5 NUCLEAR REGULATORY COMMISSION
!e g.y j wAssiucros, o. c. ros55 TO QUESTION 7 .
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%, *' **** / .
Docket No. 50-293 g gg H. R. Denton, Director, ONRR . . MEMORANDUM FOR: R. J. Mattson, Director, DSI/0NRR FROM:
SUBJECT:
GENERIC IMPLICATIONS OF THE RELEASE OF SPENT DEMINERA RES1HS FROM PILGRIM, UNIT NO. 1 i
Reference:
PHO-I-82-42/42A The release of radioactive spent resins from the Pilgrim has been Power reviewed Station, for generic t reported in PHO-I-82-42, June 11,1982, Based on infomation in ! implications in accordance with your request.on information in the docket file, and the PN and its update of June 14, 1982, on infomation obtained in telephone discussions with Region I representatives, a licensee representative, and the Operating Project Manager (DL), it is haveour bothconclusion that there generic and licensee areimplications.
- specific several related factors in this incidj items (1) through (5) below.
(1) It is probable that the resins observed and reported in the PW originally escaped from operations involved in aResins resin cleaning were operation for condensate demineralizer resins. apparently forced up a vent pipe into a ventilation exhaust duct, Vent from which the resins were transported by ventilation air flow. pipes are designed to maintain tank pressure close to atmospheric as tank levels fluctuate and gases evolve from' tank contents. Such a design provides a controlled exhaust system rather than a discharge into the building atmosphere; many such vents are present in plant designs. While it is considered good design practice to 1,nstall screens or filters in such vent lines, there were apparently no such devices in the Pilgrim vents. The Standard Review Plans 11.2 (Liquid Waste Management Systems) and 11.3 (Gaseous Waste Management Systems) and Regulatory Guide 1.143 (Radwaste System Design Guidanc do not specifically address such a design criterion. (2) It is probable that water entered the ventilation exhaust ducts along ' with the resins noted in (1), above. While it is not known if this water was significantly radioactive, the presence of the water may have been a factor Ventinlines the serving deterioration of filters liquid systems andbefilter frames should (see (3), below). designed to incorporate a device or mechanism, such as a water trap, to, prevent the flow of liquids into vent pipes discharging to ventilation exhaust ducts. Neither the applicable Standard Review Plans nor the applicable Regulatory Guide address such a design feature. pp.. S mme , h f -. -
H. R. Denton 2-ML 8 GS2 (3) The licensee considers the most probable source of the discharge of radioactively contaminated resins to the roof and ground areas of the plant to be the reactor building ventilation exhaust duct. Based on the dispersal pattern of the resins, we arrived at the same conclusion. As noted in (1) and (2), abwe, resins are presumed to have entered tank vent pipes leading to ventilation ducts, probably in the fom of a slurry. The continuous flow of wam dry air would cause the resin to dry out, leaving a residue of small beads or particles of low density, which can be carried along the duct by the ventilation exhaust air current. In the filtration plenum, air from the ventilation exhaust ducts is passed first through a fiberglass prefilter media and then through a HEPA (High Efficiency Particulate Air) filter. Air flow through the filters is horizontal
- and there is about a four-foot space (measured horizontally) between the prefilter banks and HEPA filter banks. Linear face flow velocity (design) of the prefilters is about 250 linear feet per minute, or Each HEPA filter module has a dimensional cross-section about of about 3 mph.
4 f t2and has a rated capacity, when new, of 1,000 cfm at a 1" (water) pressure drop; the face velocity for a HEPA filter is also about 250 linear feet per minute or about 3 mph. An IE Health Physics appraisal team visited Pilgrim in January and ! Feb ruary, 1980. The team's report, dated July,1980, noted that the prefilters were "disintegrating in place" (Section 4.2.3.2, page 55) but that no damage to the HEPA filters could be observed by visual inspection. This situation was apparently not corrected until the refueling outage which began in September,1981. In fairness to the licensee, though, it should be noted that the prefilter disintegration was not included as a "significant finding" by the WRC in the appraisal. While there may be extenuating circumstances which are not apparent from the IE appraisal, there appear to be no reasons why these non ESF systems could not have been taken out of seWice for replacement or repair in a more expeditious manner. O J __..,_,_.________-._,.,___.__,c.__, _
,,,,_________.__.,___m m~.,,_. . _._ .. _ _, . _- _ _ . _ .._,.,_,_.
H. R. Denton -
.JJL 8 1932 While we have not been able to detemine the exact condition of the HEPA filters at the time of their replacement in. September,1981, licensee representatives did state many of the HEPA filters were found to be damaged. It should be pointed out that no release of resins had been identified at that time and no tests were perfomed to detemine the nature or extent of leakage or damage. The staff considers that the Pilgrim occurrence has no direct implications as to the integrity of adequately tested and maintained HEPA filters in ESF filter systems but, rather, emphasizes the need for regular testing and surveillance where a specified level of perfomance is to be acMeved and maintained. The occurence is, however, a clear demonstration that plant operators cannot neglect HEPA filter systems indefinitely and then expect them to perfom as designed.
~
We note, however, that in the present regulatory climate, licensees, in general, have no compelling motivation to perfom surveillance which is not femally required of them, especially when inoperability o' a system will not lead to noncompliance. The fact that deteriorating prefilters were obsened during the Pilgrim Health Physics appraisal and that radioactive resins were found to be present in the ventilation exhaust ducts was not evidence that Technical Specification release limits or Appendix I criteria were being exceeded and, there-fore, there was no violation of regulatory requirements to initiate corrective action. The periodic testing, or replacement of non-ESF filtration system components represents an expenditure of money and manpower with little tangible benefit when only routine nomal operation is considered; in an era of tight inoney and budgetary restraints, plant managers may be hard-pressed to justify to upper levels of utility management the expenditure of even a few thousands of dollars at a very high cost-benefit ratio. j (4) Technical Specifications require periodic testing of ESF filter systems l at nearly all plants, as well as surveillance of parameters such as l pressure drop, which are indicative of system condition and perfomance. l Normal ventilation exhaust air filter systems are not ESF systems and, therefore, are not subject to Technical Specification requirements for i testing and suneillance. Non-ESF ventilation exhaust filter systems l are installed in nuclear power plant buildings to reduce releases of l airborne materi:1 to levels that satisfy the criteria of Appendix I to 10 CFR Part 50; Pilgrim, Unit 1, is only one of many plants which do not regularly inspect, check, or test their non-ESF filter systems. v._..._ , _ . _ . _ . - , , _ _ _ , _ - , . _ _ _ _ . . , _ , _ , _ . , , _ _ , , . _ , _ __.__,_.,n._, .,__,_,w.,,--en,-.__.
3% H. R. Denton l While the failure or procrastination on the part of operating plants to regularly test and assure the proper functioning of these systems may be interpreted by some parties as failing to provide maximum l protection to the environment, making such testing a fim comitment would necessitate a substantial revision in the basic HRC philosophy of plant safety and environmental protection. Comitments made by applicants in their FSAR to Regulatory Guide 1.140, "Design, Testing, and Maintenance Criteria for Nomal Ventilation Exhaust System Air - 1 Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," is the method currently used by NRR to implement design guidance and testing programs for non-ESF filter systems. Such criteria had not been established by the NRC when Pilgrim 1 was
- licensed in 1972, so it is likely that no comitment was ever made by Boston Edison to provide surveillance testing of the non-ESF filters at Pilgrim 1.
(5) The Licensee and IE (reference IE Health Physics Appraisal Report for Pilgrim, dated June 22,1980, page 54) have been aware for over two years that radioactive resin beads and fines were present in Pilgrim ventilation exhaust ducts. The same appraisal report, page 55 notes serious deficiencies in the condition of ventilation exhaust prefilters and the presence of approximately six inches of spilled radioactive (2R/hr) resins on the floor of a room in the Radwaste Building (p. 48), as well as loose contamination up to 90 mrads/hr on the floor inanediately outside that room. In view of the unique and highly visible nature of resin beads, the rather high radioactive contamination levels associated with the resin, and the knowledge that resins had been a problem in several areas of the l plant for over two years, the Licensee's statement (PN Update June 14, 1982) that the resins had probably been released prior to September 1981 seems to indicate, at best, an absence of recognition of potential problems on the part of plant management. To admit that external plant contamination of this order of magnitude nad gone unnoticed and undetected for over eight months would seem to admit to the existence of inadequacies in the Health Physics program. IE COORDINATION Our review has been coordinated with IE personnel at Bethesda, Region I, and the Resident Inspectors' office. The Radiological Safety Branch (IE) is currently reviewing completed Health Physics appraisal reports for other plants to identify any similar circumstances.to confirm the generic nature of the Pilgrim incident and support the need for issuance o
l l l H. R. Denton ' JUL 8 1982 guidance to licensees; this review has not been. completed but will be made available at a later date. ,
SUMMARY
As the result of our review of the Pilgrim, Unit 1, PNO of June 11, 1982 l (PNO-I-82-42), the staff suggests the following: (1) As a short-term action, recommend to IE that an infomation notice be issued to all operating reactors which (a) describes the Pilgrim 1 resin dispersal event, (b) requests plants to voluntarily institute a surveillance program for existing non-ESF filtration systems if . one does notiexist and (c) requests that tank vent designs be reviewed l and that, if appropriate and feasible, modifications be made to prevent
- inadvertent release of resins or liquids to the ventilation system.
NRR staf f is available to provide assistance to IE in the preparation i of such a circular. (2) As a longer tem action, revise Regulatory Guide 1.143, "Design Guidance for Radioactive Waste Management Systems, Structures, and Components installed in Light-Water-Cooled Huclear Power Plants," and Standard Review Plan 11.2, "Liquid Waste Management Systems," to include design j guidance and acceptance criteria which address (a) the incorporation I of filters or screen 5. in the design of vents from tanks which may contain resins, and (b) the incorporation of prwisions into the vent design such as filters traps or check valves to prevent or minimize t l the flow of liquids through vent lines while pertnitting pressure equaliz-ation within the tank. k d a R. J. att , Dir or Division of Systems ntegration Of fice of Nuclear Reactor Regulation l
~ " ' ~ ' ~ ' ~ '" = , . . . , . _ , . , _ . _ _ . _ _
JUL 8 1982 cc: E. Case -
- D. Eisenhut ,
S. Hanauer l G. Laines T. Novak W. Houston W. Gamill D. Yassallo F. Congel
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L. Hulman R. Bangart - C. Willis R. Capra L. Cunningham s K. Eccleston P. Stoddart d i N N O I
. . . . - . ~ - . , - . . - - ._ , . , . . _ _ _ . . . . - - . . - . - - - . - . - - - - , . - . , - . ~ . - . . . - - . _ .
UNITED STATES ENCLOSURE 6
,4 pa usg),, NUCLEAR REGULATORY coMMisstoN TO OllESTION 7 g w Assmotow, o. c. nosss
. E. . ;3 <-
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~'
APR 191983
/
MEMORANDUM FOR: Xarl V. Seyfrit, Chief AEOD/T307 Reactor Operations Analysis Branch Office for Analysis and Evaluation ' of Operational Data THRU: Stuart D. Rubin, lead Engineer i Reactor Systems 4 Reactor Operations Analysis Branch FRG4: John L. Pellet Reactor Systems 4 ' Reactor Operations Analysis Branch .
SUBJECT:
TECHNICAL REVIEW REPORT ON PILGRIM 1 RESIN MlGRATION i Enclosed find the technir81 review report titled"Condensate Demineralizer
- Resin Migration Through the Plant Vent and Standby Gas Treatment System." .
This report concludes that no additional AEOD/ROAS involvement is necessary I
) for this event. !
b 'ON d - i 1 John L. Pellet ) Reactor Systems 4 , Reactor Operations Analysis Branch l 1 A 4nH~ > , , + vjqJW6 yL -
- 1 !
\
9*
AEOD TECHNICAL REVIEh' REPORT *
~
UNIT: Pilgrim 1 TR REPORT NO.: Atoo/T307
- 50-293 DATE: April 19, 19B3 DOCKE.T:
LICENSEE: Boston Edison Company EVALUATOR / CONTACT: J. Pellet ( NSSS/AE: General Electric /Bechtel
SUBJECT:
CONDENSATE DEMINERALI2ER RESIN MIGRATION THROUGH THE PLANT VENT AND THE STANDBY GAS TREATMENT SYSTEM . EVEh7 DATE: June 11, 1982
SUMMARY
, 1 This report reviews the safety significance of the June 1982 discovery at Pilgrim that demineralizer resins had migrated throughout the plant contam-inated exhaust vent to external plant areas inside the protected area fenc-ing. Also, sufficient resin had migrated through the reactor building .
ventilation system to block proper operation of the Standby Gas Treatment System (SBGTS). "eferences are cited which show that resin migration into the ventilation system and SBGTS had occurred at least three years previously. This report finds that the event was of minimal safety significance and con-cludes that current NRC ef forts are adequate without additional AEOD involve- - ment.- , , DISCUSSION . Plant & Status Pilgrim I was in steady state power operation on June 11, 1982 while perfoming a surveillance instruction (SI) on the SBGTS. Occurrence-Cause & Effect1 l The SBGTS f ailed its routine SI due to low flow. The low flow was caused ' by carryover of. resin beads from the condensate demineralizer vent piping to the reactor building ventilation system and contaminated exhaust vent and from there to the SBGTS. This carryover occurred during backwashing of the demineralizer. .Backwashing with air and water resulted in resin fines, particulates, and some resin beads being entrained in the air / water backwash.
- An air scrubber was installed during initial startup to prevent resin migration 1nto the ventilation system. However, it did not perfem as expected since installation. As a result, substantial resin migrated to the radwaste and ventilation systems over a considerable time period.
Af ter this event, contaminated resin beads were discovered outside of the Less plant buildin2s (but not offsite) as well as inside the vent system. than 70 cubic feet of resin was removed from the ventilation system and less than 1/2 of a cubic foot was found inside the protected area. Root cause of the substantial resin migration appears to be inadec.uate design of the scrubber intended to preclude such migration. j ,
,1 eai1 t ritr*
7 n.JWJW7 W J N ^g /
'inis cocument supports ongoing. AE00 and HRC activities and does not represent the position or requirements of the responsible NRC program office.
Histor At least two cases of resin intrusion into the SSGTS have been previously reported2 .3 since June,1979. This indicates that resin intrusion into the ventilation system and SBGTS has been a recognized problem at Pilgrim for several years without adequate resolution. However, prior to the June 11,1982 event there was no evidence of contLmination outside of the plant buildings. Consequences The consequences of this event may be broken down into three categories:
- 1) offsite release, 2) personnel exposure, and (3) system performance or '
avail ability. The resin migration problem produced no evidence of offsite
~
release during this review. However, the resin migration clearly has resulted in added equipment contamination and substantial cleanup efforts by plant personnel over a period of several years, but this review found no indication of unacceptable personnel exposure. From a system viewpoint, this event demonstrates the potential for failure in a nonsafety system to . act as a cmmon cause initiator affecting multiple trains of a safety system ' (in this case SBGTS). This potential is mitigated because failure is as a r
. result of flow restriction due to resin buildup and is therefore very slow j with respect to the test interval (i.e., only two f ailures over the last i three years). Also, even though one train of 55GTS was inoperable due to low air flow, the train was capable of perfoming at a-reduced level.
In summary, the resin migration produced minimal actual consequences in 1 the three areas of concern. {, - Corrective Actions - The licensee actions to preclude further resin migration into the vent system may be divided into short-tem and long-term efforts. The immediate actions by the licensee to remove existing resin and preclude additional migration were set out in Confimative Action Letter No. 62-19,. Additionally, the licensee disconnected the ventilation system from the poorly functioning gas scrubber and rerouted the scrubber dischargeHowever, (liquid, air, and resin) to the Reactor Building Equipment Sump. l the equipment sump was not intended for eitner the quantity of air / water ' mixture or the entrained , resins produced by demineralizer backwashing. This resulted in sump discharge to the HPCI toom during demineralizer backwash. Due to a loose cap on a floor drain, approximately 12 inches i l of water accumulated in the 5 RRR pump room as weil at in the HPCI rocm, The licensee ' Resin contamination was also evident in the HPCI room *. corrected this problem by securing the leaking floor drain and admin-istratively requiring low sump leve.1 prior to desineralizer backwash. The above details introduce considerable uncertainty as to the long-term ' efficacy of the corrective actions implemented by the licensee thus f ar. The licensee is currently studying potential lon;-tem corrective actions . and can be expacted t6 impienent such actions v.t.en they are dettmined. - The'URC Resident Inspector is following this subject and can bt: expected to require an adequate resolution based on his past efforts. I
. - . . - _ - - . , - - - l
o .
-3 FINDINGS Findings for this investigation were:
- 1) _ Resin migration through the ventilation system can produce a common mode failure of both trains of SBGTS.
- 2) The safety significance of this event is minimal due to the ' slow propogation rate and limited actual consequences of the resin migration.
. 3) Corrective actions by,the licensee are adequate at present.
CONCLUSIONS The safety significance of this event is relatively minor given the radio-logical release and system performance effects previously discussed. The personnel exposure effects may be more significant, especially since this has evidently been a problem for over three years. However, this review produced no evidence of excess personnel exposure. Given the limited signi- . ficance discussed above, followup and resolution of this event by the resident inspector appears to be adequate. At present there is no need for. additional AEOD involvement en this event. However, this type of; common mode failure is l i potentially generic, depending on plant specific arrangement of 'demineralizer
. vents, SEGIS, and reactor building ventilation. l REFERENCES . .
l '. LER 82-019/03L-0 on Pilgrim unit 1.
- 2. LER 79-020/03L-0 on Pilgrim Unit 1.
3.. IE Inspection No. 50-293/B2-20 l l
. 4. Confirmative Action Letter 82-19
- 5. IE Inspection No. 50-293/82-30 .
l 1 a . l 1 e
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l !' EliCLOSURE 7 TO QUESTION 7 _ Event Evaluation Sheet , Information subject Source Initial Receiver Date - l~< 1u in.w 4j7s</$L b , o , i. e4 Sys&r2s" as c. (fivo.sr<.s) Q2 '12) ~ Regional Contact Licensee / Facility Type / Location _ [ ^" '- *~ y,-. W r a.
. P \ Lcs - /8 un /RE
,d<'sc< & n s & n ,,,. , n Event Sum,ary: t/n/s2., sp Lr pU ~~op'sl} cws v-l 4nq,L>t. cl & p / f".,wec.k <M J ,y.,,L r.s, /
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t# 1 W 5/ ft>5 To l L su - -=- -Pet % ' * * ^' N T Ts~@~n~cD ~6 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ W #'" l Further Action Assioned to - (- equired_ No (AssignCodef)* Yes ___ ____ Followup Actions /Results_ cv clt- , u , I5 i..m.a sins vs3 Luohk. % d.1y *vf.2 v a r.w .n.a g y p i i f. c u.v.. .,. I .l . ' . . . p hU
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b p. b'.r ~ sv I , PN Is 3 s c el . Item Closed by/Date ( . Concurrence /Da te_ ) (Sect. Chief) * \
l oc$ so: 50293-820611 Date: June 11. 1982 fgl N ua T11 *;3J.0j CF INE 47 OR UNJSt AL CCCURRENCE--PNO-I-32-42 . . . 1h:: p alis.1 u ry i>otsfient ion esnstia utan EARt Y notice af events' of POSSIBLE safety or r ui f: increr.t v piff.:anca. The infomation is as initially received without verifi- j cacon er avs' us:* on 4,v1 is basica)1v all that is k,nown by the Region I staff on this c' te. g Pitt, rim Nuclent ?ower Station faglpy Plyno tth, Vast,ac7usetts L.leensee Emergency Classification: DN $0 293 _, Notification of Unusual Event
. . Annyt -
Site Area Eeergency ) General Emergency _,,f,_,Not Appitcable e .. .. l Su!*i;1 : SELTASE OF 59G7 REstN A.a.iro s mr.el; 1300T.4.bsequent on hne 11. 1982 spar t risin was found ,c.n, th,,e yoygd,tpar the ut rveys identified contomination of the roofs of the T ,i bis e Wi Nf.19 Ta bit.e. f ear. tor, Xf-ir s and Re-Tube Buildings. Contamination was also found on the Contamination levels renged from 20-30,000 ger uid ot). in t's nte s entrolle d Are5f. dx /1K 2l- bi ta mu.ir.w. c. ratamina". ion of i.p tc ICO,iX;0 dpm/100 cm'. Gama isotopic a u Irt.is c f the esin itancified primarily long lived radionuclides (Co-60, Cs-137 I C,. ly snc dn-54). I
*is contstr.inari an was IJa.ntified off-site or in stern crains. All personnel are being f.1 st: 4 prior to exiti10 tiit site and no personnel contaminatien has been identified.
T s. isin mh .1,we been released throu'gh the reactor buildin0 vont duct which exhausts The licensee has found t tio s.ami. hex sb aaofelovaticn rusin in of Standby the arpmximately Gas 100 ff., System Treatment inlet plenum. The a yN is,tely 9: fc Three rsdiaticn specialists have been 2,t rese t' t u. n s 11 is t aine Investigated. d'tplicacc w the site to t' valuate the radiologicel aspects of the occurrpocat.. 4.cis in:, era.t .s .ege :ted cu2 to public interest in tho facility. The licensee is c rsic' aran; isshi.i; a pt en r11 ease. The NRC does not plan to issue a press release b i1 et11 re+ond to me,lia 't.quiries. 1ha Como.1 wealth of Massachusetts has been lat o 7at.t. .. .. Iliis l'N u, arwnt as of 4 45 P.M., June 11, 1982. W T.sCT- Eltaster 3 runner . 4 45-1235 488-12:25 -
%5 h!B11Dh 4 5;. -
M4BE Phillips ~ EN Willste ~ Mail: ADM:DMS l , :l.ai n a.i Vla . hno E00 ~ ~ ~ ~' N AR IE NMS5 : 00T:Trans. On1 !
'or a Gili.it.ky PA 01A RES '~ "~~ ~ ~
l
'ons. Ahaarna PA -
' AE00 t et . Ree ri.s ELil i tLR$ Air' Rights INPO ~~
l iECY SP N S AC,~
- - - , .* -.., ,.. ~--~ ---
- - ~ ~
l j, , 2DR As sicai O f f t:e> TMI Resident Section
~~ ~
RI Rerident Office ,, __,,_,,,,,, U cent.es: (heactor ticenseus) h gien Fun 83 M..y * ? st', snn ? o n - m m l1:e r u re z m r c jP.
,_ 2
l I [ . . . . ,_. _ f a ? n, - U 50293-8.20611. _ W t- ~ri [ Date: DCS Mo.:Jdne 14. 1982 - e f- a .! L ,1 : l y tE -PHO-I-82 42A ,, l hip!ERf ignFICATIch ij,' EVENT OR %!SUAL t 00C1JRRD M:- . . t .' e of events of POSSIBLE safety or ! Ms preiit' nary. notifica1'ica constitutas EARLT notic fritially received without verifi- l Alle titarest signif' car ice. The information-is as n by the Region I 's' tam o'n'
' ion ei evalstiun, atd ,f a sasically all thak is Ar ce - t l
1s date. ' ** " - - - - I C Pilgrim Nuclear r Station . Licw see Emergency Classification: l
' j lh:C111ts:
Notification of Unusual Event
,,' 31Pboouth.
SS.293 Massapusetts Alert t
., i
' - Site Area Emergency i i
- - . General Emergency
~.
__x Not Applicable F RELEASE OF IPENT RESIN (UPDATE PMO-t-82-42); yg , l l surveys of selected areas of the brveys of *.te cnt;re !.ite wi*hin the erotected area t identification of the spent resin / lcevee contr111ed area $re made within 3 hours of hieoo. T1e heensee s ensite surveys identified twg contaminated tion of thepavement areas which Turbine, Administration. ins tarticaded avi poster Surveys confirwd contamiN i Otan ted 03f-4as and Re-ibe Building roofs. The Reac< 4r Buildi.ng, Roof we,s ..fqund. f o be ~ frte of con amination. Die licensee's offsite survey. No contam' included surveys of cars, parking nation was identified. Routine i
. hts, shc rr ro.it, and se rity access arees. Nothir;g bimfmental air samales overing the period June 1 15: 1982 were counted.
H.14.s141 was id mti' led. i. cause of the size and This weightiof wan confirmed the resins, by airnosamples offsite airbornecollected Aleare c f 2.e beads appetrs to have occurrect. ihr11p cletn u? of the co fta 11r.a:ed pt.venent areas which the when Retcounted indicated background
; tor Buiding Vent. Preif min-a v tre ident!'ica. ion of b sasiples of stem fralq; residueresins have only on roof-tops been counted kihh no undef contamination A duct surveillance identified.
program All e.catoninati tentitatrtn ducts have been va'cuumedic: een. has b6en as .abitsned to id entify any additional resimmum lation. f ,
'. !l resin entered the ventilation iu c t s .,from the . .co nd en.a te.. d emi n e r-TE: liew.sw W11 aves the In addition, eizur 33 stear Aring .esir backwashinp via the Cation Regeneration Tank Vent.
in fecm slefective r,cndcr. sata iemineralizar vent valves may have also been released or to their repair c'urfng the September 1981 -March ],982 refueling outage. The resin l pears to t.svo been relaisert from the Reactor Building? Ventilation Exhaust System which
. fit above thn reactor h il t ng roof. nrior to jthe .repqir of defective filters in this l systam in Septamber 1981. .. ,;
I
#be licat.see has susperdec l all transfer oportati'ons which could result in further resin hisasas to vent,ilt4 tion di ets and has initiated addit <cdalation environmental sampling. The Specialists througftout the liccuste's 1.ctions wen mnitored by three Region I Rad <
w> eked, Region I will i ue a Confirwatory Action Let@r to address planned licensee
'corrwthe ectiens. Tr e icensee is continuird to review the source and cause to detemine khat pesaM.nt c'orrective etion will be needed. The; Resident Inspectors are closely Volluming iteensee action ; coricerning this event.
8 2 . l4ndia Inte tst has occurrdd. The licensee has resoondedmedia The NRC will respond to to media inouf'inouiries ries but btit'doesdbes not not plan p;an to isn:e a petss r:1gse. - so tusue a prose,.reletse. a. .. .. . . i This PN is current as of 1:00 a.m. . June 14. 1982. j i enn/n e ? in . n
; m 1 9 . u .; }j
EtiCLOSURE 3 TO n!HESTInn 7
,' [*
i SSINS No.: 6835 IN 82-43 UNITED STATES NUCLEAR REGULATORY C0$911SS10N OFFICE OF INSPECT 10H AND ENFORCEMENT WASHINGTON, D. C. 20555 Noveeber 16, 1982 i IE INFORMATION NOTICE NO. 82-43: DEFICIENCIES IN LWR AIR FILTRATI0tt/ YENTILATION SYSTEMS Addressees: All nuclear power reactor facilities holding an operating license (OL) or l construction pemit (CP).
Purpose:
l J This information notice is provided as notifier. tion of events that had actual or potential radiological impact on the plant environs. It is expected that l recipients will review the information for applicability to their facilities. l No specific action or response is required. Description of Circumstances: I Within the past 2-1/2 years, air filtration / ventilation systems at five I facilities were found to have serious deficiencies, ranging from overloaded i prefilters to evidence of a wetted high-efficiency particulate air (HEPA) ) filter bank, to penetration of HEPA filter banks by substantive quantities of radioactive resin beads. Deficiencies occurred in both safety-related and non-safety related systems, in June 1982, radioactive spent resin was found on the grounds and roof areas at Pilgrim 1. Princioal radionuclides were Co-60, Cs-137, Cs-134, and Mn-54; contamination ranged from 20,000 dpm/100 cm' to 100,000 dpm/100 cat. The contamination penetrated damaged filters in a non-safety-grade HEPA filter plenum. The degraded condition of these filters was not detected in a timely unner because of a lack of surveillance or testing of the filtration system. l The HEPA filter failure occurred possibly as an end result of a combination of high dust loadings and mechanical dange resulting from the impact of disintegrating prefilters, as well as the probable warping or distortion of HEPA filter frames under prolonged exposure to water and high humidity. l In Decester 1980, the SGTS trains at Brunswick 1 were found to be operating at close to 1001 humidity, and condensation was observed on the interior walls. i Regulatory Guide 1.52 reconnends operation at humidity of 70% or less; operation at high humidity is known to cause substantial degradation of the iodine-retention capacity of charcoal adsorbers. Also, in Deceeter 1980, both filter trains in the turbine building filter system at Brunswick were found to be operating with the upstream HEPA differential pressure gauges offscale high. Also, in the turbine building filter system, 431 of the upstream HEPA filters were improperly installed.
4 y IN 82-43 i Noverber 16, 1982 Page 2 of 2 In August 1980, filters and charcoal adsorbers in the Surry 1 process vent exhaust air treatment system were detennined to have been half submerged in water, and the HEPA filters were caked with dust. No pressure drop instru-mentation was provided across the filter banks to ascertain their state of loading. Also, in August 1980, pressure drop gauges across the HEPA filter banks in the ventilation exhaust treatment system of the auxiliary building at Surry 1 exceeded 5 inches, which is offscale high; this condition had existed since May 1980. In May 1980, the nonnal containment building exhaust filters at Turkey Point were found to be overloaded with dust to such an extent that the filter medium was separated from its frame in more than 50% of the filters. This apparently allowed radioactive contamination resulting from explosive plugging of steam generator tubes to be transported to the Southeast sector of the plant site. In March 1980, it was determined that HEPA filters in the Big Rock Point offgas and chemistry laboratory exhaust treatment systems were not being tested for leakage in place. No records were maintained of pressure differential across the laboratory HEPA filters which had not been replaced for at least five years. In each case described above, licensees initiated programs and procedures to correct the deficiencies and to preYent or minimize their potential for reoccurrence. Air treatment systems which incorporate filtration or adsorption media are provided to reduce the potential release of radioactive materials to the environs. In order to function as designed, such systems should be installed. tested, and maintained to a degree consistent with their intended function. Guidance on installation, maintenance, and testing programs, of a degree and nature which have been demonstrated to ensurt proper system functioning, is provided in Regulatory Guides 1.52 and 1.140. No written response to this infonnation notice is required. If you need addi-tional information about this matter, please contact the Regional Administrator of the appropriate NRC Regional Office or this office, utalle Me 4, i
- Edward L. Jordan, Director l Division of Engineering and !
Quality Assurance l Office of Inspection and Enforcement l l Technical Contacts: L. J. Cunningham, IE 301-492-8073 P. G. Stoddart, NRR 301-492-7633 l
u
$Nsori 7 Pilgrim Nuclear Power Station Radioactive Effluent and Waste Disposal Report including ,
RadiologicalImpact on Humans January 1 through June 30,1982 l By: Nuclear Operations Support Department Environmental and Radiological Health and Safety Group
. Date: September 1,1982 Bo,eton Edison Company D DC O!hhhh93 ggs:
R pgg
PILGRIM NUCLEAR POWER STATION RADI0 ACTIVE EFFLUENT AND WASTE DISPOSAL REPORT INCLUDING RADIOLOGICAL IMPACT ON HUP %NS l l JANUARY 1 THROUGH JUNE 30, 1982 I i l l Prepared by: Ib_ r _--
\
Christine E. Bownan Sr. Radiological Engineer Approved by: _, Thomas L. Sowdon Environmental and Radiological Health and Safety Group Leader Date of Submittal: Septeeber 1.1982 l l 1 r - -- - - - - - - - . , - . - , , . , , - . -, - - - - - - - - - -
TABLE OF CONTENTS Page Seefion i
- 1. Introduction and Sisenary 1
- 2. Effluent. Waste Disposal and Wind Data 41
- 3. Off-Si+.e Doses Resulting from Radioactive Liquid Effluents Off-Site Doses Resulting from Radioactive Gaseous Effluents 46 4.
67 l
- 5. Off-Site Doses from Direct Radiation i
i LIST OF TABLES l Page l Table 2 Supplemental Infonnation 3 1A Gaseous Effluents - Sumation of All Releases 4 18 Gaseous Effluents - Elevated Release 5 1C Gaseous Effluents -Ground Level Release Liquid Effluents - Sumation of All Releases 6 2A 7 2B Liquid Effluents Solid Waste and Irradiated Fuel Shipments 8 3 4A-1 Distribution of Wind Directions and Speeds - 33 Ft. Level 9 ' of 16qFt. Tower 4A-2 Distribution of Wind Directions and Speeds - 160 Ft. Level 25 I of 160Ft. Tower 42 3.2-1 January-June 1982 Liquid Releasa Maximisn Individual Doses from all PatNays for Adults (MREM) 43 3.2-2 January-June 1982 Liquid Release Maximum Individual Doses from all PatNays for Teenagers (MREM) 44
' 2 3 January-June 1982 Liquid Release V4ximisn Individual Doses from all Pathways for Children (MREM) ,
3.3-1 Population Ooses Rasulting from the January-June 1982 45 Liquid Effluents iii
Page Table. LIST OF TABLES (Cont.) 47 4.1-1 Un# pleated Relative Concentration per Unit Emission for Reactor Building Vent for January-March 1982 Depleted Relative Concentrations per Unit Emission for 48 4:1-2 Reactor Building Vent for January-Narch 1982 Relative Deposition Concentrations per Unit Emission for 49 4.1-3 Reactor Building Vent for January-March 1982 Undepleted Relative Concentrations per Unit Emission for 50 4.1-4 Main Stack for January-March 1982 51 4.1-5 Depleted Relative Concentrations per Unit Emission for ~ Main Stack for January-March 1982 Relative Desposition Concentrations per Unit Emission for 52 4.1-6 Main Stack for January-March 1982 Undepleted Relative Concentrations per Unit Emission for 53 4.1-7 Reactor Building Vent for April-June 1982 Depleted Relative Concentration per Unit Emission for 54 4.1-8 Reactor Building Vent for April-June 1982 Reittive Deposition Concentrations per Unit Emission for 55 4.1-9 Reactor Building Vent for April-Jure 1982 4.1-10 Undepleted Relative Concentrations per Unit Emission for 56 Main Stack for April-June 1982 4.1-11 Depleted Relative Concentrations per Unit Emission for 57 Main Stack for April-June 1982 4.1-12 Relative Deposition Concentrations per Unit Emission for 58 Main Stack for April-June 1982 Maximm Individual Locations and Pathways 59 4.2-1 January-June 1982 Gaseous Release Maximum Individual Doses 60
- 4. 2- 2 fmm all Pathways for Adults (EEM)
January-June 1982 Gaseous Release Maximum Individual Doses 61 4.2-3 from all Pathways for Teenagers (REM) January-June 1982 Gaseous Release Maxime Individual Doses 62
- 4. 2- 4 from all Pathways for Children (MREM)
January-June 1982 Gaseous Release Maxime Individual Doses 63 4.2-5 from all Pathways for Infants (iiREM) . iv
- - , - , - . . - - ,n,----- - - - - - , , . , . - - - - - - , , - - . . - - . . . , ,--,,,,-.-.._,,--,.,._,n, , , . . . , , - - , - - . - ,
Page LIST OF TABLES (Cont.1 163 64 4.2-6 January-June 1982 Gaseous Release Maximun Individual Doses 0.5 Miles SE 65
- 4. 3-1 Population Distribution 66
- 4. 3-2 Population Doses Via Major PatNays Resulting from Gaseous Effluents during January-June 1982 6
Y
t
- 1. INTRODUCTION AND
SUMMARY
This report is issued for the peMod January-June 1982 in accordance with NRC Regulatory feuide 1.21 "Measuring. Evaluating and Reporting Radioactivity in Solid Wastas and Releases of Radicartive Materials in Liquid and Gaseous Effluents from Light-Water Cooled Nuclear Power Plants" (Rev.1) . The inforration supplied includss actual effluent releases, radioactive waste and meteorological data; doses from liquid releases, doses from gaseous releues and direct gansna radia-tion doses.
- 2. EFRUEE, WASTE DISPOSAL AND WIND DATA Radioactive liquid and gaseous releases <. wind speed data together with measurement errtrs and solid wasta disposal information are given in Tables 1 A.18.1C. 2A. 25, 3. 4A-1. 4A-2. and supplemental information section in the standard Regulatory Guik 1.21 fonnat.
I l 1
. 1 I
EFPLUENT AND WASTE DlWCSAL El!MtANNUAL REPORT Sussiementalitdermmen January - June 1982 p aw,_5eim Nedar P sneen ue ces u _
. arveier usets Qs + QY
- a. Femam sne wrueimia seats 'S I 0.25/ E 0.10/E
- h. kdvin2C1/ Quarter
- f. Penamies. heirape w 9,i 13(1.8E4Qs +1.8E5QY)W 1
- d. Lged efnwett.10C1/yuarta r
- . insum m Perm-se co ee,mme Pisem;e time WPra used m deterfreney allismable rettane rates we omstneratums Fasam and ac1:vaine pants e CFR se a.
- b. kernet Apposes 3
- c. Panavistes. half 4 ries >m days Totte []
- d. bqual efnwntt M . 3 e t X l91 JCVat; as east. IeCTR 34 Appenea S. Table D
- 3. AverseeEmere pnmde it. swran entr MS =0. 324 : RBV=0.e 503 iL vi ihe redammi.le musiure m ewae, oi ras.m and wnam um
- 4. Ideenwersients and Appresuriosissa of Total RadisectMey Pnmde the northuds used tu rneawre or setenunwe the itssi rainwittet) in e(Dents and IN trettnah und in detercine todamwkde oweresta a.
- a. Fasam and unstne panes- ) g,y
- h. lodines: f' Pin otalet-
){ W t.
- d. Lewd e f ownes'
- l. Roset Rc4menes Pn ede ele Ivikmsg eformaine retaing tu heich reiesars of radia1:w tretensis m laqud aM pescius efnuenas
- s. L& quad
- i. we ser of nacw r,6eawr 121
- 2. Tuial tuvo pered for bach eveemes: 192.92 hrs
- 3. Womuan t3 prrud for a teich e*me - 7.75hr$
- d. A.efapt time permd tot he Ji reisens 1.59 hrs
- 5. Weimum tirse penod f.m a heich reisew - 0.25 hrs A Averspr ureem fk= durms periods of retenee of efownt sto a thierwg 6sreem 1.90E+5GPM
- b. Cassee Odes AppEntie)
L Ah mai sehmes a. b MDG 1 2 i
5 TABLE 1 A EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT GASEOUS EFFLUENTS. SUMMATION OF ALL RELEASES January - June 1982 Quarter Querior Est. Total uaN 1 2 Enor. % A. Fission and activation gases Q - 3.55E+3 2.50E +1 l
- 1. Total nleese
- 2. Average release rate for period sC1/sec
- 4.52E+2
- 3. Percent of Technical Specifiestice limit % - 6.92E-2 B. Iodines
- 1. TotalIML% 131 Q -
3.97E-3 2.54E+1 l
- 2. Average release rete for period sC1/sec 5.05E-4
- 3. Percent of Techrdeal Specifiestion limit I % - 1.992-1 i C. Partleulates
- 1. Particulatas with half. lives > 8 days a <3.68E-4 4. 26E- 3 3.05E+1 l
- 2. Average release rate for period sC/see < 4. 7X- 5 5.42E-4
- 3. Percent of Technical Specincetion timit n < 8.39E-3 6.98E-2
- 4. Gross alpha radioactivity Q < 4.52E-7 G.61 E-7 D. Tritium Q 2.34E0 5.92E0 3.20E+1 l
- 1. Tota! release
- 2. Average release rate for per.od sC1/sec 3.01E-1 7.52E-1
- 3. Percent of Technical Speci5 cation limit % -
l A. i
l i I TABLE 15 l EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT ( 1982) GASEOUS EFFLUENTS - ELEVATED RELEASE January - June 1982 j CONTINUOUS MODE SATCH M00f uart omerw l beror l omrw l ourw l l l l mmw l l
- 1. Flesion gases 0 - 1.37E-2 irwton45 -
krypton 45m Q 2.9 3E+2 0
- 6.55E+1 krypton 47 Q - 3.62E+2 l bypton48 !
0 - 2. 28E+ 3 2enon.133 - Q 2.61E+2 senon 135 Q 4 6.06EM Dnon 135m menon 138 Q - (2.38E+1 l
- l xenoc 131m Q -
l Q xenon 137 menon 133m Q - 4.28E+1 f 0 - 3. 33E+ 3 l l Total for period l 2 Iodines l l
!~Rne-131 0 -
2.5X-3 Q
- I.90E-3
- iodine-133 l Q - 46.55E-3 iodine-135 Total for period 0 - 41. 70E- 2
- 3. Partjeuham 0 4 6.32E-7 5.16E-4 strontium 49 strontium 90 Q 4 6.26E 8 5.50E-6 coef e .134 0 Q 4.1.04E-5 1.14E-5 cesium 137 0 1.5 7E- 3 barium tanthanum.140 chromium 41 Q Q 5.90E-6 2.90E-6 -.
manganese-54 cobalt 48 Q Leon 59 Q Q 4 7.86E-5 3.00 E-5 cobalt 40 nac45 0 sitconium-niobium 95 0 cerium 141 Q omrium 144 Q ruthenium 103 Q ruthenium 106 Q
1 1 l 1 TABLE 1C t l EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1982l
- GASEOUS EFFLUENTS. GROUND LEVEL RELEASE :
January - June 1982 CONTINUOUS M00E BATCH MODE ou,w j u=t j h j omnw l o.nw l l j l News men =d
- 1. F'ission gases krypton 85 0 - 1.01E-5 krypton 45m Q 2.4)E+1 krypton 47 Q -
2.51E+0 l C - 4.55E+1 krypton 48 0 - 4.19E+1 zenon 133 C1
- 1.07E+ 2 senon.135 -
sanon 135m Q - senon 138 Q - Total for period 0 - 2.22E+2 2 Iodines 0 - 1. 44 E- 3 Hlne 131 Q - 6.50E-3 lodine 133 ladine 135 0 - 41.02E-2 Total for penod Q - 41.81E-2
- 3. Particulates Q 1.64E-5 1 46 E-3 strontium 89 strontium 90 Q 4.76E-7 1. 44E-6 essium 134 Q 1.17E-6 Q 2.42E-5 3.67E-5 cesium 137 Q 3.95E-4 barium lanthanum.140 0 1.08E-5 5.88E 6 manganese-54 cobalt SS Q -
teon 59 Q EMt40
- O 2.16E-4 2.27E-4 sine 45 Q sitconium niobium 95 Q oertum 141 Q ,
rut' enium 103 Q ruthenium 106 0 ,,
. .s .
r , - -
TABLE 2A
> EFFLUENT AND WASTE DISPOSAL SEMlANNUAL REPORT (1983 LIQUlO EFFLUENTS SUMMATION OF ALL RELEASES January - June 1982 unn Y A. Fission and activation prodnets
- 1. Total release (not including tritium, Q 5.72E-1 1.44E-1 3.00E+1 nobk m, w alpw
- 2. Average duuted concentracon pQlm) 7.58E-8 8.91E-8 during period
- 3. Percent of appucabw umit % 5.72E0 1.44E0
- 8. Tritium Q 5.26E0 1.99 E-1 3.00E+1 l
- 1. Total release
- 2. Average duuted concentration gC/ml 8.19 E-7 1.05E-7 during period _
- 3. Percent of applicabk unit % 8.19 E0 1.05E0 C. Dissolved and entrained gases
- 1. Total release Q * -
j
- 2. Average duuted concentration sci /ml , ,
during period
- 3. Pareens of applicable limit %
D. Gross alpha radioactivity [ 1. Total release l Q I41.44E-4 141.73E-5 l 4.00E+1 l E. Volume of weste released (prior hters 1.61E6 1.10E5 2.00E+1 to dilution) F. Volume of dilution water used liters 6.42E9 1.90E9 2.00E+1 during period 6-
TABLE 25 EFFLUENT AND WASTE DISPOSAL SEMI ANNUAL REPORT (1982) LIQUID EFFLUENTS January - June 1982 CONTINUOUS hA005 BATCH MODE Querise Querest Quarter Quarter Nwedidos Reinemed Unet Q 6. 70E-4 1.89E-3 strontium 89 0 4.17E-4 1.65E-4 strontium 90 cesium 134 Q l.46E-2 7.42E-4 Q 1.08E-1 6.60E- 3~ cesium 137 0 - 2.25E-6 6edme 131 a 2.54E-3 8. 23E-4 cobalt.58 Q 2.44E-1 7.00E 2 cobalt-60 Q 4.27E-5 3.06E 6 tron 59 Q 4. 28 E- 3 1.20E-3 nac45 - ' 0 2.61E 2 1.01E 2 Jnanganese 54 0 - 1.20E-5 chromium 51 , siteonimum-niobium 95 0 5.16E-4 6.74E-4 molybdenum 99- *
- technetium 99m Q Q - 4.96E-5 barium lanthanum 140 .
eerium 141 Q 1.65E-5 -
- 2.70E-6 iodine-133 - Q {
a - 1.75E-5 eerium.144 eQver 110m Q iron 46 0 1.47E-1 2.4 3E-2 l 2.40E-2 l 2.72E-2 j l urddentised l Q l l Q 5. 72E-1 1.44E-1 Total for petiod (above) senon.133 0 - zenon136 Q _, 7
TABLE 3 EFFLUENT AND WASTE DISPOSAL SEMI-ANNUAL REPORT (1982) SOLID WASTE AND !RRA0!ATED FUEL SHIPMENTS JANUARY - JUNE 1982 A. SOLID WASTE SHIPPED 0FF SITE FOR BURIAL OR DISPOSAL. (Not irradiated fuel.) UNIT 6M EST. WAL
- 1. TYPE OF WASTE PERIOD ERROR 1 Spent resins, filter slud9es, m3 97.299 N/A l
- a. Ci 123.60353 N/A )
evaporator bottoms, etc. l Dry compressible wasta, contaminated m3 15 N/A
- b. .
equiprrent, etc. !
- c. Irradiated co@onents, control m3 NONE N/A I rods , e tc . Ci l
i
- d. Other (Describe) m3 NOME N/A !
Miscellaneous im-level wasta Ci l
- 2. ESTIMTE OF MJOR NUCLIDE COPPOSITION. (By Type of Wasta) l l
1 E(Curies) I
- a. Spent Resins. Filter 5:10 .522 .64564
$1udoes. Evao. Sottoms. Sre9 19.972 24.68618 Fes5 12 607 15.6945a Di a toma tacus Ea rth . E tc .
Cs134 4.156 5.13671 Cs137 26.127 32.54062 Cc58 1.220 1.50773 1 2 54 2.712 3.35228 Zn65 .450 .55669 C060 31.633 39.09916 La-140 .019 .02323 Ba.9an .005 .00671 1-1 31 .004 .0049' ___ Cr 51 .283 ""*
. >8
~
TUTAL5 100.000 123.60353 _, _ 8A .
s 1 E(Curies) , 50.24 5.36260
- b. Dry Comoressible Wasta Co60 .81467 Contamina ted Equipment Co58 7.63 22.48 2.39956 Cs137 fi.75 .72011 Cs' 34 '
.18635 FeUS . 75 1 .14 .12171 Fe59
.12 .0132B Sr89
.01 .00027 Sr90 In65 .Z3 .02a88 9.65 1.03030 Mn54 100.00 10.67373 TOTALS
- c. M/A
- d. M/A
- 3. SOLID WASTE DISPOSITION Mode of Transportation _ Destina tion Nurter of Shionunts.
Richland, Wash, 20 Tractor Trailer Tractor Trailer Earnwell . S.C. 32 B. IRRADIATED FUEL SHIPMENTS (Disposition) Nwter of Shiopents Mode of Transportation Destination N/A N/A NONE e
Ei! CLOSURE 10
. - TO QUESTTON 7
'c, PILGRIM NUCLEAR POWER STATION Radioactive Effluent and Waste Disposal Report including Radiologicalimpact on Humans 1
July 1 through December 31,1982 f l BY: NUCLEAR OPERATIONS SUPPORT DEPARTMENT ENVIRONMENTAL AND RADIOLOGICAL HEALTH AND SAFETY GROUP ; Date: March 1,1983 4 BOSTON EDISON COMPANY [889;88848888h PDR
-_ . - . - _ _ - _ _ _ _ _ _ _ _ _ . ~ - _ - __-...- - _, - __.-. .__ _ _ ..-___
l ) PILGRIM NUCLEAR POWER STATION 4 RADI0 ACTIVE EFFLUENT AND WASTE DISPOSAL REPORT INCLUDING RADIOLOGICAL IMPACT ON HUMANS JULY 1 THROUGH OECEMBER 31, 1982 4 i Prepared By: $In1 I ba-- Christine E. Bewman , Senior Radiological Engineer Approved By: , r=7 Thomas L. Sowdon ! Environmental Radiological 4 Health and Safety Group Leader Date of Submittal: March 1,1983 4 i I
TABLE OF CONTENTS Paoe Section
- 1. Introduction and Sumary 1
- 2. Effluent. Waste Disposal and Wind Data 1
- 3. Off-Site Doses Resulting From Radioactive Liquid Effluents 41
- 4. Off-Site Doses Resulting From Radioactive Gaseous Effluents 46 -
- 5. Off-Site Ooses From Direct Radiation 68 LIST OF TABLES Table Paoe Supplemental Inforration 2 3
1A Gaseous Effluents - Sumation Bf All Releases 4 18 Gasicus Effluents - Elevated Release 5 1C Gaseous Effluents - Ground Level Release L 2A Liquid Effluents - Summation of All Releases 6 2B Liquid Effluents 7 3 Solid Waste and Irradiated Fuel Shipments 8 4A-1 Distribution of Wind Directions and Speeds - 33 ft. Level 9 j of 160 ft. Tower 4A-2 Distribution of Wind Directions and Speeds - 160 f t Level 25 of 160 f t. Tower l 3.2-1 July-Deceder 1982 Liquid Release Maximum Individual Doses 42 from all Pai.hways for Adults (MREM) 3.2-2 July-December 1982 Liquid Release Maximum Individual Doses 43 from all Pathways for Teenagers (MREM) i 3.2-3 July-December 1982 Liquid Release Maximum Individual Doses 44 l from all Pathways for Children (MREM) 3.3-1 Population Doses Resulting from the July-December 1982 45 i Liquid Effluents l l iii l
- . t LIST OF TABLES (cont.) ,
Pace Table 4.1-1 Undepleted Relative Concentrations per Unit Emission for 47 Reactor Building Vent for July-September 1982 , 4.1-2 Depleted Relative concentrations per Unit Emission for 48 Reactor Building Vent for July-September 1982 j 4.1-3 Relative Deposition Concentrations per Unit Emission for Reactor Building Vent for July-September 1982 49
~
4.1-4 Undepleted Relative Concentrations per Unit Emission for Main Stack for July-September 1982 50 4.1-5 Depleted Relative Concentrations per Unit Emission for i 51 Main Stack for July-September 1982 4.1-6 Relative Deposition Concentrations per Unit Emission for 52 Main Stack for July-September 1982 4.1-7 undepleted Relative Concentrations per Unit Emission for Reactor Building vent for Octe:;er-December 1982 53 4.1-8 Dep1.eted Relative Concentrationsler Unit Emission for
-Reactor Building Vent for October-December 1982 54 ,
4.1 9 Relative Deposition Concentrations per Unit Emission for Reactor Building vent for October-December 1982 55 , 4.1-10 Undepleted Relative Concentrations per Unit Emission for 56 Main Stack for October-December 1982 4.1-11 Depleted Relative Concentrations per Unit Emission for 57 j . Main Stack for October-December 1982 , 4 4.1-12 Relative Deposition Concentrations per Unit Emission for 58 Main Star.k for October-December 1982 4.2-1 Maximum Individual Locations and Pathways 59 4.2-2 July December 1982 Gaseous Release Maximum Indi.idual Doses from all Pathways for Adults (MREM) 60 4 4.2-3 July-December 1982 Gaseous Release Maximum Individual Doses 61 from all Pathways for Teenagers (MREM) 4.2-4 July-December 1982 Gaseous Release Maximum Individual Doses 62 from all Pathways for Children (MREM) ! 4.2-5 July-December 1982 Gaseous Release Maximum Individual Doses from all Pathways for Infants (MREM) 63 I iv 'i
- _ _ _ . - - _ - _ - - - < -,---,-y..-,._.--.,,.--.,--,,m,,-. -
, .y . _. - - - , - - , , - ,.. . - - ,-,y9,,--- ,-,-.,-r---,._. - . ., r.-
t h LIST OF TABLES (cont.) t Table Page s 4.2-6 July-December 1982 Gaseous Release Maximum Individual Ooses i
- 0.6 Miles ESE 64 l 4
4.3-1 Population Distribution 65 l 4.3-2 Population Ooses Via Major Pathways Resulting from Gaseous Effluents during July-December 1982 66 h k a l . l 1 4 I i i 4 i V i I l
I 9
- 1. INTRODUCTION AND
SUMMARY
S This report is issued for the period July-December 1982 in accordance with NRC Regulatory Guide 1.21 "Measuring. Evaluating and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water Cooled Nuclear Power Plants" (Rev. 1). The information supplied includes actual effluent releases, radioactive waste and meteorological data; doses from liquid releases, doses from gaseous releases and direct gamma radiation doses.
- 2. EFFLUENT WASTE DISPOSAL AND WIND DATA Radioactive liquid and gaseous releases, wind speed data together with measurement errors and solid waste disposal information are given in :
Tables IA, 18. 1C2 2A, 28, 3, 4A-1, 4A-2, and supplemental information l section in the standard Regulatory Guide 1.21 format.
]
l I 1 l
)
i i I 2 1
+ EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPOQT Supolemental Informetson . , July-December 1982 Ucense, DPR45 Fedhty _ Phlsvvi Nucles Posee Sunon r i, amisim umiii
- s. Fiuion and a.inaimn sases OS OV = 5. 1 0.25/I-- + 0.10/E '
- h. er heiaes 2Cirariisvain. hafaan >qua rter13(1.8E4Qs + l .8E5Qv)51
- e. =ani
- d. L9=J enteea'5 1001 per quarter 2 Maumum Permiss ble Concenirstion Piovide the MPCs used m determinieg atto 4tste re!ene rates or umentraine Fissain and astiv4in:n gases 10 CFR to 4
- h. h. dines Appendia 5
- s. Particulaies. half 4nes >= Jo s Table ll J. UquiJ efiiuena s H 3. t X 101.Cl/mi, 1 all rest.10 CFR 20. Appendia B. Table II
). Average Energv Provide ite average energy tU ut ibe rada.nustae meuore m iete4w.ui tinavi anJ 4stivativri g4ws. it 4pph 4efe. E .1 we, MS = 0.30480.287:
Measurements andRBV = Approumations 391 c ovity of Toreg Assee&O.494 (3rd & 4th quarter) 4 Proede ibe nie h.4s used h. measwee of arp.oum46e 4he loial radnia.iivii) in tilbents and the meib.4s uwJ to determine radnino.hJe somresinon
- a. Fnsain and 4 iiration pses Geu b lodines 3,,g si PJ rtig wfales d LqwJ eiduenis 4thnis
$. Batch Relesws Pio.4e itv tvilo.ing inform 4uon retauer to eat.h releases si eadnia tive maienals m hqaid and gaseow, e<t'weet.
- a. Uqwd I %wmNr of bai.h releaies 77
- Tot al tune perioJ eoe hai6h rele4ws 87.48 hrs
- 3. Maumum time pern d for a eaich release - 4.08 hrs 4 A.erage ume persed tur bai h seteases 1.14 hrs
- 3. Minimum time pens 4 few a hatch reitase - 0 33
- 6. Aier4pc stream (b. Jwneg penods of release eni unto aoi*eTn, no ing stream hrs 3.05E+5 GM b Cameow (Not Appuesble)
.. ~,m.1 me.es
- b. N0fle 2 1 j
I TABLE 1 A EFFLUENT AND WASTE DISPOSAL SEMlANNUAL REPORT GASEOUS EFFLUENTS SUMMATION OF ALL RELEASES July-Decerrber 1982 Quarter Quarter Est. Total Unit (3) (4) trror.% , A. Fission and activation gases Ci .%1.07E+4 % 5.19E+3 2.49E+1 l
- 1. Totai relene
- 2. Average releue rate for period uCi/sec < 1.35E+3 % 6.53E+2
- 3. Percent of Technical Specifiestion limit % < 1.77E-1 % 8.25E-2 B. lodines
- 1. Total lodine.131 C1 1.03E-2 '
9.32E-3 2.51E+1 ]
- 2. Average releue rate for period uCWsee 1.30E-3 1.17E-3
- 3. Percent of Technical Specification limit % 5.15E-1 4.66E-1 C. Particulates
- 1. Particulates with half lives > 8 days Ci 8.20E-3 8.01E-3 3.03E+1 l
- 2. Aversge releue rate for period uCi/sec 1.03E-3 1.01E-3
- 3. Percent of Technical Specification 11mit % !
9.67E-2 8.72E-2
- 4. Gross alpha radioactinty Ci l45.14E-7 < 4.50E-7 D. Tritium ci
- 1. Totai releue 4.90E0 5.93E0 3.30E+1 l
- 2. Average releue rate for period uCi/sec 6.16E-1 7.46E-1
- 3. Percent of Technical Specification limit % ,
l j b 3 l
TABLE 18 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (19821 GASEOUS EFFLUENTS - ELEVATED RELEASE July-December 1982 ;
.- - l CONTINUOUS MODE BATCH MODE l No.we. m.w l umi l ou.ner I ov.ner I c a.n.r ! ou.n.r j (3) (4)
- 1. Fission gases krypton 85 Cl 1.62E-2 1.60E-2 krypton 85m Ci 7 69E+2 5 47E+2 krypton 87 Ci < 1.87E+2 < 4.58E+1 krypton 88 Ci 8.99E+2 4.99E+2 xenon.133 Ci 4.51E+3 3.07E+3 menon 135 C1 3.73E+3 7.36E+2 xenon 135m Cl d.1.54E+1 4 9.26E0 senon 138 Ci < 3.75E+1 4 3.90E+1 zenon 131m C1
. menon 137 C1 zenon.133m C1 1.30E+2 8.49E+1 Total for period Ci <. 1. 03 E +4 _ 6.03E+3
- 2. lodines l
iodine 131 Cl 4.66E-3 6.53E-3 } i iodine 133 Ci 1.68E-2 2.24E-2 iodine 135 Ci < l.22E-2 < 1.48E-2 Total for eened Ci < 3.37E-2 < 4.37E-2 , 3. Particulates 1 i strontium 89 Ci ' l.62E-3 2.78E-3 strontium 90 C1 1.73E-5 1.83E-5 I cesium 134 Ci 8.15E-6 2.61E-6 eesium 137 Ci 7.38E-5 5.76E-5 bartum tanthanum.140 C1 3.55E-3 2.68E-3 chromium 51 C1 - manganese-54 Cl 1.28E-5 3.65E-6 I cobalt 58 C1 - 2 09E-6 tron 59 Ci - l cobalt 60 Ci 1.55E-4 3.97E-5 : ! .me65
' Ci - -
! l "irconium
. niobium 95 C1 - -
eenum 141 Ci j eenum 144 C1
- 1.53E 5 j ! ruthenium 103 Ci nathenium 106 Ct 2.70E-5 -
4 , 1 i I
a TABLE 1C EFFLUENT AND WASTE DISPOSAL SEMI ANNUAL REPORT (1982 ) GASEOUS EFFLUENTS . GROUND LEVEL RELEASE July-December 1962 CONTINUOUS MODE BATCH MODE Nuclides Rowsed Unit Overter Quarter Ouarter Owarter j (3) (4)
- 1. Flasion sues krypton 85 Ci < l.49E-5 5.03E-6 krypton 8'm Ci < 3.46E+1 1. 21 E+1 krypton 87 Q < 9.16E0 < 4.07E0 krypton 88 Ci < 1.55E+1 2.43E+1 zenon.133 Cl 1.41E+2 5.99E+1 ,
renon.135 Cl 1.86E+2 5.86E+1 senon.135m Ci - . tenon.138 C1 Total for period Q < 3.86E+2 < 1.59E+2 l
- 2. lodines iodine.131 Q ; AAr 1 2.79E-3 iodine 133 Q 2.63E-2 1.18E-2 lodine.135 Ci 4.26E-2 2.10E-2
~
Total for penod Ci 7 46E*2 3.56E-2
- 3. Particulates I strontium 89 Q 1.29E-3 1.53E-3 I
, strontium 90 C1 2.55E-6 2.53E-6 i cesium.134 Ci 1.89E-6 4.46E-6 ! cesium.137 Ci 6.64E-5 2.14E-5 barium lanthanum.140 Ci manganese.54 Ci I . Z b t. - 5 1.31E-6 3,74E.6 I cobdt.58 Ci - iron.59 Ci - . cobalt 60 Q 1.29E-4 5.90E-5 zine 65 Ci 1 zirconium. niobium 95 Cl . - eenum.141 Ci . . j ruthenium.103 Ci . . 2.60E-5 nathenium 106 Ci - 5
TABLE 2A EFFLUENT AND WASTE DISPOSAL SEMI ANNUAL REPORT (1983 LIQUID EFFLUENTS SUMMATION OF ALL RELEASES JULY-December 1982 d5 Net Ow#$r a Est. Totat Unit Error. % A. Fission and activation products
- 1. Total releue (not including tritium, Ci
- 1 -l 98M noble gues, or alpha)
- 2. Average diluted concentration gCi/mi 7.39E-9 6.65E-8 j during period
- 3. Percent of applicable limit % 3.09E-1 1.25E0 B. Tritium _.
j 1. Totai release Ci 8.29E-4 4.55E-1 3. 00 E+1
- 2. Average diluted concentration uCi/ml 2.42 E-7 1.98E-10 during period
- 3. Percent of applicable limit % 1.98E-3 2.42E0 C. Dissolved and entrained gues
~
- 1. Total releue C1
- 5.39E-3 3.98E+1 l
- 2. Average diluted concentration uCilmi
- 2'O ~
during period
- 3. Percent of applicable limit % - -
D. Gross alpha radioactivity l 1. Total relene l Ci l46.60E-6 141.65E-5 l 4.01E+1 l E. Volume of wute released (prior liters 8.4 7 E+4 2.01E+5 2.00E+1 to dilution) F. Volume of dilution water used titers 1.88E+9 2. 00 E+1 4.18E+9 during period 6
_. . _- _~ . . _ i . i ! l 4 . j TA8LE 2B t i EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1984 ' LIQUID EFFLUENTS l July-December 1982 l I CONTINUOUS MCOE BATCH MODE i Nuclides Roleesed Umt OwN Quaher Quarter Overtoe ! strontium 89 Ci 1.64E-5 2.10E-5 strontium 90 C1 4.70E-5 7.78E-5 cesium 134 Ci 3.30E-4 7.05E-4
; cesium 137 C1 3.73E-3 9.65E-3 -
c 4 iodine 131 Ci 5.87E-6 4.12E-5 cobdt 58 C1 4.42E-5 1.96E-3 t cobalt 60 Ci 8.67E-3 3.66E-2 ! iron 59 Ci 3749E-6 5. 30 E-4 I unc65 - Ci 5.09 E-5 5.37E-5 manganese 54 Ci 6.49E-4 3.74E-3 J chromium 51 Ci 4.02E-5 6.57E-3 l
- l !
rirconimum niobium 95 Ci - 1.21E-6 I I
! molybdenum 99 Cl
.- 5.71E-5 ;
technetium 99m barium lanthanum 140 Ci 1.03E-6 4.3SE 5 t ! cenum 141 Ci 2.14 E-6 1.10E 4 - todine 133 Ci - 3.04E-6 l cerium 144 Ci - - I silver 110m C1 - 8.01E 4 ~ tron 55 Ci 1.28E-2 2.41 E-2
)
i j [_ unidentified l Ci l 4.49E-3 l3.95E-2 l l l i i Total for period (above) Ci 3.09E-2 l 1.25E,1 1 : - - . j menon 133 Ci - 2.18E-? _ xenon 135 Ci - 3.21E.J i 1 4 i
TABLE 3 ETTLUENT AND VASTE DISPOSAL SEMI-ANhTAL REPORT (1982)
- SOLID WASTE AND IRRADIATTD TUEL SHIPMINTS
- JULY - DECEMBER 1982 t A. SOLID WASTE SHIPPED OTT SIII TOR BURIAL OR DISPOSAL. (not irradiated fuel) l UNIT 6 MONTH EST. TOTAL
- 1. IYPE OF VASTI PERIOD ERROR % ,
4
- a. Spent resins, filter sludges, m3 99.007 N/A
! evaporator bottoms, etc. Ci 819.10 N/A Dry compressible vaste, m3 547.666 N/A b.
contaminated equipeent, etc. Ci 5.14564 N/A
- c. Irradiated components, m3 none N/A control rods, etc. Ci none N/A
- d. Other (describe) m3 none N/A f N/A Miscellaneous icv-level vaste Ci none
- 2. ESTIMATE OF MAJOR STCLIDE COMPOSITION. (by type of vaste) 3 E(Curie:)
- a. Spent Resins. Filter Co-6 0 41.324 339.4ag20 I
t Sludges. Evaporator Co -58 3.864 31.65107 j I Bettems etc. Cs-137 13.426 109.97069 i Cs-134 1.489 12.19371 Te-55 11.164 99.44832 l Te-59 .597 4.89055 I-131 .464 3.79925 I-133 .070 t .57668 l l 14-140 .220 1.80569 Ba-140 .019 .15592
' Sr-89 15.478 126.78505 Sr-90 .345 2.82477 1
Sr-91 .003 .02146 Tc-99m . CA O .32557 1 Zn-65 .723 5.92615 Mn-54 4.614 37.7374v
- 8A -
i ) i 1
_. _ _ _ _ . ._ . . . _ _ ._ =-. _ . _ . . _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ i 1 : - i (by type of vaste)
- 2. ESTIMATE OF MAJOR NUCLIDE COMPOSITION.
- j CONTINVED
% E(Curie s) ;
l
.002 .01495 !
i Seent Resins. Filter $1udte s, ht-9 5 a. Evap. Sottond. Diatomateous Cr-51 6.090 49.88606 ! i 3 Earth, etc. As-110m ( .001 .00641 l Co-141 .030 .24916 l l continued Ru-103 .014 .11290 ; i i " Sr-92 .001 .00691
.010 .08267 I Sb-124 Xe-133 < .001 .00034 ! ! .03266 Xe-135 .004 i Mo-99 .007 .05629 [
I .
- . i TOTAL
- 100.000 819.10682 l t
% E(Curies)
~
f [ l Co-60 17.46 .89843 l j b. Dry compressible Vasts. ! Contaminated Equipment Co-58 6.32 .32546 i Cs-137 6.04 .31058
' cs-134 1.65 .08565 .
Te-39 1.17 .06038 f I-13: , 2.74 .14116 } ! .19341 f ' Ba-140 3.76 2n-6 5 .86 04410 - i l 3.19 .1711A i j Mn-54 Cr-51 56.60 2.91179 l l - w1AL: 100.000 5.14564 ! 1 t 6 i !
- c. N/A .
{< ! 1
- d. N/A '
t j 3. SOLID WASTE DISPOSITION ' Number of Shipments Mode of Transportation De stina tion l 37 Tractor Trailer Barnwell. S.C. ! Tractor Trailer Richland. Wash, t j 2 i I
- 4. IRRADIATED TUEL $HIPMENTS (Disposition) ;
I Number of Shipments Mode of Transportation Destination i N/A I none N/A , i i
\
t i - 8s - 1 I I t i
b 4 QUESTION 8. In recent years, Boston Edison has had unsatisfactory ratings i in the area of fire protection. I would like to know if Pilgrim is now in full compliance with fire protection requirements? Are all barriers, fire doors and penetration j seals repaired and capable of passing required testing? Are fire watches still required in certain areas of the plant? Ilow many fire watches are still needed? Will the NRC require 1 Edison to complete the upgrading of the entire fire protection system prior to allowing restart? How many maintenance requests are still outstanding in the area of fire protection? Please also coment on the condition of the Halon system in the computer room at the plant and the smoke detectors over the spent fuel pool, t ANSWER. 1 Pilgrim is either in compliance or will be in compliance with its fire protection requirements prior to restart. l i l During the last one and one-half to two years, Boston Edison Company has made significant improvements in their entire fire protection program. Additional personnel with extensive experience in nuclear power plant fire protection , i j have been hired. Realignment of responsibilities and authority among these
! licensee personnel have strengthened the entire fire protection program and i I i
i i i l
QUESTf0N 80 (Con?' o$ 2 : l provided a higher 4' 2w , and continuity of effort that has resulted in substan, 4 1'r- sent the program. This is evidenced by the l methodology and thorougni sss exhibited in identifying and correcting deficiencies. k One activity of the additional licensee fire protection personnel described above was the licensee has performed a reevaluation of plant fire protection features, comparing those features against NRC requirements and guidance, in an effort to determine (a) the level of actual compliance, and (b) the adequacy i of the features provided to prevent unacceptable fire damage. 4 During the course of this reevaluation the licensee found several cases where they did not literally comply with the NRC requirements or specific , coninitments they had made earlier. The licensee, however, provided justification to demonstrate adequate protection against unacceptable fire damage and on that basis, asked for exemptions from those requirements. In ! l i most cases the staff granted the exemptions. In those cases where the staff ; did not agree with the justification provided, the licensee made modifications so as to be in compliance.
; Because of the more or less constant activity at operating plants, temporary ) changes, repairs and, modifications, may result in a particular condition that ) is not in compliance. These situations are contemplated by the licensee and j provisions are in place to assist in identifying the situation beforehand, J
1 providing interim protection measures (such as fire watches) and maintaining administrative control of the situation to assure that the out-of-compliance
)
condition is corrected, i ; I l i
QllESTf0N 8. (Continued) 3 The licensee has indicated that all modifications and work associated with upgrading required fire barriers, fire doors and penetration seals has been completed. The licensee has committed to having all of the necessary document-ation concerning the above work completed prior to plant startup. , Fire watches continue to be used in some areas at pilgrim as well as most operating plants. At the beginning of the present outage approximately 18 months ago, eight persons per shift were assigned full time responsibility for continuous or roving fire watches covering approximately 180 individual deficiencies. As of March 17, 1988, no continuous fire watches are required. Two persons per shift are assigned roving as fire watches covering 41 separate deficiencies throughout the entire plant. Of those 41 deficiencies, 25 are related to fire barriers, 15 are related to maintenance activities, and one is related specifically to activities pertaining to the outage. Some minor upgrading to the f' ire protection systems may remain at the time Pilgrim restarts. However, those modifications yet to be completed will have been identified and the schedules for completion will have been reviewed for I acceptability by the staff. l One hundred and sixty-one maintenance requests were still outstanding in the area of fire protection on March 17, 1988. However, this number by itself does not give an accurate picture of the Pilgrim fire protection maintenance program. On January 5, 1987 there were 260 open maintenance requests related to fire protection. Since January 1,1987, approximately 1,480 new fire protectien-related maintenance requests have been generated and approximately 1,580 have been closed. l l
QUESTION 80 (Continued) 3 You also asked for our coments on the condition of the Halon System in the computer room, and smoke detectors over the spent fuel pool. A computer located in a small room adjacent to the Cable Spreading Room is being phased out. The room is protected by an operable automatic Halon fire suppression system. A new plant computer has been installed next to the Technical Support Center and the primary fire protection is provided by a sprinkler system with secondary protection provided by an automatic Halon fire suppression system. Both of , these systems are operable. Six smoke detectors are located over the Spent Fuel Pool in the ventilation system exhaust ducts. Four of the six detectors have already been tested during this current plant outage. The other two are scheduled for testing prior to plant startup.
QUESTION 9. How many automatic and manual scrams have occurred at Pfigrim since the plant became operational? What is the annual industry-wide average? ANSWER. Table 1 provides data on unplanned automatic and manual scrams during operational modes (criticality to 100% power) for Pilgrim from 1984 through 1987 compiled from licensee event reports submitted pursuant to 10 CFR Part 50.72 and 10 CFR Part 50.73. The comparable industry average rates are also provided in Table 1. Prior to 1984, reactor scrams were not directly reportable to the NRC (Pilgrim entered comercial service December 1,1972).
Enclosure:
Table of Unplanned Scrams When Critical for Pilgrim and Irdustry
. . .. . _ - _ - _ _ _ _ = . .- - .
i i i Enclosure to Questir:n 9 i. ! Table 1 i ! l, 9 } l l Unplanned Scrams When Critical for Pfigrim and Industry 1
+ 1984 - 1987
- l i ;
'l I e 1984* 1985 1986 1987** j ] 1 i e b 1 i Pilgrim ; 1 ; I i ) Autoratic 0 4 4 0
- i. l i
) [ Manual 0 0 0 0 I I i .i
)
l l ! i 1 i ] i ) Industry Average l ' I Automatic 5.4 5.0 4.0 3.2 j l l Manual 0.6 0.5 0.5 0.6 ! 1 I l ! 1 j
- Pilgrim critical hours for 1984 = 170. )
i l "Pilgrim critical hours for 1987 = 0. l 2 !
)
I
QUESTION 10. How many "Unusual Events" and how many "Alerts" have been declared at Pilgrini since 1972? Please describe and oive the date of each report. How does this compare to the industry-wide average? ANSWER. l The NRC did not use the terms "unusual events" and "alerts" until 1980 and did not established them as reportable categories in our regulations until 1983. Our computer records of notifications to the NRC Operations Center show that Pilgrim has declared 12 Unusual Events and no Alerts since 1983. Of the 12 Unusual Events, 2 were caused by fires in nonsafety related equipment, and 1 was due to a potentially contaminated individual being transferred offsite for medical treatn ant. The remainder were attriouted to safety system in-operability, which necessitated shutdown of the plant in accordance with the plant's Technical Specifications. Two tables are enclosed - the first compares the number of unusual events at Pilgrim since 1983 with the industry average per year; and the second provides descriptive data and the date for each unusual event at plants.
Enclosure:
Tables of Unusual Events at Pilgrim Nuclear Station
OVESTION 10. (Continued) 2 A comparison of Pilgrim Unusual Events versus the industry average follows: Industry Unusual Licensed Industry Pilgrim Unusual Year Events Units Average Events
- 1982 -
1983 205 85 2.4 0 1984 224 91 2.0 1 1985 312 98 3.2 5 1986 209 104 2.0 5 1987 231 109 2.1 0 +1988 - 5 Year Total IT 7 TI
*This table was prepared from data contained in computerized data base from August 1982 to the present. For comparison purposes, incomplete data for 1982 and 1988 are not shown. However, Pilgrim did report Unusual Events (a fire in a face mask fitting machine) on August 18, 1982 and on February 11, 1988 (a fire in the machine shop). Pilgrim also had one Alert on June 3, 1982 relating to a withdrawn incore detector resulting in abnormal radiation levels.
This event lasted approximately 2 hours. Pilgrim had no other Alerts from 1983 to 1987; however, Alerts have been reported from other licensed facilities.
QUESTION 10. (Continued) 3 Enclosure to Question 10 Unusual Events at Pilgrim Nuclear Station August 1983 to Present Event Description , 4/26/84 Potentially contaminated man taken to hospital. 5/16/85 2 safety system trains inoperable. 05/23/85 2 safety system trains inoperable. 09/20/85 2 safety system trains inoperable. 10/15/85 2 safety system trains inoperable. 11/04/85 Residual Heat Removal safety train A inoperable. 01/04/86 2 of 8 Main Steam Isolation Valves fail closure time test. 01/09/86 Fire in line to hydrogen storage tariks. 02/11/86 Low pressure coolant injection inoperable. 02/14/86 2 safety system trains inoperable. 04/11/86 Loss of containment integrity. 02/11/88 Fire in machine shop. l l l l l l I l i j
i l l 00EST10N 11. How many violations of NRC regulations have occurred at Pilgrim since it began operation? What is the industry-wide average? ANSWER. The NRC does not itaintain industry wide statistics on the total numbers of violations per plant. In order to provide this requested data for the Pilgrim facilties, a review of inspection report data was performed. Our review indicated that Pilgrim was cited approximately 425 times for violations or deviations since the plant began operation in June, 1972 through the end of 1987. This number however, does not reflect whether the citations involved individual or multiple violations, whether the citations were subsequently withdrawn, or the severity level of the vio-lations. Moreover, enforcement history is only one of a variety of factors NRC considers in assessing licensee performance. 1 i l
I QUESTION 12. There have been a number of allegations concerning the illegal dumping of radioactive waste on Boston Edison property. Concerns have also been raised over Edison's use of the town dump for disposal of radioactive material. Would you please describe what monitoring the NRC conducts or requires on materials and wasta leaving the Pilgrim site. Has the NRC or the licensee performed tests on Edison property and at the town dump to ensure that there are no elevated levels of radiation at areas suspected of containing radioactive waste? Where and when were tests conducted? What were the results? ANSWER. The NRC staff does not itself monitor materials and waste leaving the Pilgrim site. The licensee is required to monitor all items containing or contaminated with radioactivity that leave the site and there are several facility procedures that provide specific guidance and instructions to plant nealth physics workers regarding this activity. All radioactive wastes that are sent to sites specifically intended for burial must meet federal regulations for radiation dose rate and contamination levels as well as special requirements of the burial sites. NRC performs routine inspections of the radioactive transportation area to ensure that licensees are conforming to these regulatory requirements. Further, onsite i materials that have the potential of being contaminated and are being shipped i offsite are surveyed prior to being shipped. The licensee is not allowed to ] dispose of contaminated objects in non-radwaste facilities without obtaining a special variance required by in 10 CFR Part 20.302(a). BECo has not applied for l l l l l l
QUESTION 12. (Continued) 2 these variances. To our knowledge, no contaminated objects have been disposed of in the town dump or in other public facilities not specifically intended for contaminated objects. The NRC received allegations that contaminated shrubs had been removed from the site and improperly disposed of on BEco property in 1987. NRC inspectors determined that appropriate surveys were performed, measurements were within established limits and properly recorded prior to offsite disposal. An NRC inspector accompanied by the licensee collected clippings from the shrubs which were disposed of offsite. The clippings were independently analyzed by the NRC. Only one sample had detectable levels when we used sensitive laboratory instru-ments but was not detectable using standard survey meters. The contamination levels were lower than typical soil background levels and they posed no health hazard (see pages 12 - 13 of the enclosed Inspection Report 50-293/87-57, cated March 11,1988,p.12). NRC has not performed surveys for contamination of the town dump or at other BECo properties and does not routinely perform contamination surveys of this type. As stated in the Inspection Report, the inspectors reviewed the licensee's program for release of material from the site and concluded that it was adequate. I
Enclosure:
Inspection Report dated 3/11/88 l
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UNITED STATES NUuEAR REGULATORY COMMISSION Enclosure to Question 1 j* '"' j 2 REGION I 475 ALLENDALE ROAD
.f KING OF PRUS$1 A, PENNsYLV ANI A 194o6 MAR 141998 S:ket No. 50-293 7.as.on Edison Company ATTN: Mr. Ralph G. Bird Senior Vice Preridr.r.t - Nuclear 800 Boylston Street Boston, Massachusetts 02199 Gentlemen:
[m
Subject:
Region I Inspection Report No. 50-293/87-57 > V This refers to the routine safety inspection (50-293/87-57) cenducted by Messrs. C. Warren, J. Lyash and T. Kim of this office on December 7,1957 to January 19,19SS at the Pilgrim Nuclear Poner Statien, Plymouth, Massachusetts. Areas examined during this inspection are described in the NRC Region I Inspection Report which is enclosed with this letter. Based on the results of this inspection, it appears that one of your activities related to high radiation area access control was not conducted in full com-pliance with NRC requirements, as set forth in the Notice of Violation enclosed herewith as Appendix A. The problem was identified by your staff. However, a Notice of Violation is being issued because effective corrective actions apparently have not been taken for previous problems with high radiation area access control. In addition to following the instructions of Appendix A in preparing the required response, please include those actions you intend to take to preclude recurrence of this problem by insuring that your corrective actions are effective and lasting. Two significant integrated plant tests were successfully executed during the inspection period. Preplanning and control of these activities was generally strong. We also cbserved that increased management involvement in assuring effective problem followup has resulted in substantial improvement. Equipment i failures identified as a result of an unanticipated safety system actuation however, indicate the need for stronger post-work test practicos and a thorough i power ascension test program. The response directed by this letter and the accompanying Notice are not suoject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511. en . m m . , t /- L ' 0 b y c4 L y) q l .
Boston Ecison Company 2 g, ; ggg Your cooperation with us in this matter is appreciated. Sincerely, 1 [A m mue. . ns, Jeputy Director Divisinn nf React.nr Prniactc
Enclosures:
- 1. Appendix A, Notice of Violation
- 2. NRC Region I Inspection Report No. 50-293/87-57 cc w/entis:
R. Barrett, Nuclear Operations Manager B. McIntyre, Chairman, Department of Public Utilities Chairman, Plymouth Board of Selectmen Chairman, Ouxbury Board of Selectmen Plymouth Civil Defense Director J. Keyes, Boston Edison Regulatory Af f airs and Programs E. Robinson, Nuclear Information Manager R. Swanson, Nuclear Engineering Department Manager The Honorable Edward J. Markey The Honorable Edward P. Kirby The Honorable Peter V. Forman ' S. Pollard, Secretary of Energy Resources P. Agnes, Assistant Secretary of Public Safety, Commonwealth of Massachusetts R. Shimshak, MASSPIRG Public Document Room (POR) local Public Document Room (LPOR) Nuclear Safety Information Center (NSIC) NRC Resident Inspector Commonwesith of Massachusetts (2) bec w/encis: l Region I Docket Room (with concurrences) j W. Russell, RA M. Perkins, ORMA (w/o enc 15) j R. Blough, ORP L. Doerflein, ORP R. Bores, DRSS
- 5. Collins, ORP C. Anderson, DRS
- 0. Mcdonald, LPM, NRR T. Chandrasekaran, SPLB, NRR M. Callahan, OCA J. Nickerson
APPENDIX A NOTICE OF VIOLATION Boston Ediser. Company Docket No. 50-293 Pilgrim Nuclear Power Station License No. OPR-35 As a result of the inspection conducted on December 7, 1987 to January 19, 1988, and in accordance with the NRC Enforcement Policy (10 CFR 2, Appendix C), the following violation was identified. Three previous Notices of Violation dated March 13, March 23, 1987, and April 28, 1987 were issued for problems related to the control of Locked High Radiation Areas. It is evident that corrective actions taken in response to these Notices of Violation have not been effective in precluding recurrence. The Station Technical Specification 6.11, "Radiation Protection Program," requires that "procedures for pers vnel radiation protection shall be prepared consistent with the requirements of 10 CFR 20 and shall be l approved, maintained and adhered to for all operations involving personnel radiation exposure." The Station Procedure 6.1-012, "Access to High Radiation Areas," requires in part that the areas controlled under this procedure remain locked or guarded at all times. Contrary to the above, on December 15, 1987, December 27, 1987, and on January 8, 1988, doors to the areas being controlled as Locked High Radiation Areas were found to be unlocked and unattended, in violation of the Station Procedure 6.1-012. This is a Severity Level IV 'liolation (Supplement I). j Pursuant to the provisions of :) CFR 2.201, Boston Edison Company is hereby required to submit to this offici within thirty days from the receipt of the letter which transmitted this Notice, a written statement or explanation in reply, including: (1) the corrective steps which have been taken and the results achieved; (2) corrective steps which will be taken to avoid further violations; and (3) the date when full compliance will be achieved. Where good cause is shown, consideration will be given to extending this response time. i t
V. S. NUCLEAR REGULATORY COMMISSION REGION I Docket / Report No. 50-293/87-57 Licensee: Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 Facility: Pilgrim Nuclear Power Station Location: Plymouth, Massachusetts Dates: December 7, 1987 - January 19, 1988 Inspectors: C. Warren, Senior Resident Inspector J. Lyash, Resident Inspector T. Kim, Resident Inspector Approved By: - U /l- S I A. Randy inough, Chief Date Reactor Projects Section No. 3B Areas Inspected: Routine resident inspection of plant operations, radiation protection, physical security, plant events, maintenance, surveillance, outage activities, and reports to the NRC. The inspection consisted of 350 hours of direct inspection. Principal licensee management representatives contacted are listed in Attachment I. Observations made by the NRC Region I, Regional Ad.inistrctor during a tour on December 8, 1987 are documented in Attachment II of this report. A copy of Attachment II was provided to licensee management for follewup. t Results: Violation: Repeated occurrences of locked high radiation area doors being left open and unattended were identified by the licensee. Problems with high radia-tion area access control have been previously identified and were the subject of violations during inspections 50-293/87-03 and 50-293/87-11. Corrective actions taken in response to these findings have not prevented their recurrence. (Section 3.b, VIO 87-57-01) Unresolved Item: The licensee identified that two reactor vessel level gauges were incorrectly installed. A licensee investigation is currently ongoing to determine the cause and to assess the adequacy of post installation test. (Section 4.d, UNR 87-57-02) nnson, A iOG DOWJMAP'"' fh
Inspection Results (Continued) 2 Concerns:
- 1. The licensee experienced safety related equipment malfunctions upon receiving a spurious reactor scram signal on January 17, 1988. (Section 4.d)
- 2. Inadequate procedures and planning of surveillance tests resulted in un-necessary engineered safety feature actuations. (Section 3.a)
- 3. Poor preplanning and control of maintenance was noted during an electrical relay replacement. A similar problem was the subject of a violation dur-ing inspection 50-293/87-50. (Section 4.c) 4 Weak identification and tracking of lifted leads and jumpers led to a water spill in the high pressure coolant injection system room during the integrated leak rate test. (Section 6.0)
- 5. The prelube pump for the "B" emergency diesel generator failed to restart during a surveillance test. An identical failure occurred during a loss of of f site power event on November 12, 1987. Licensee followup appeared adequate but the failure root cause has not been identified. (Section 3.b)
- 6. The inspectors evaluated the erosion of construction dirt into wetlands area, The inspector's independent survey of the area, and the licensee's analyses indicate that the level of activity does not represent a health or safety concern. However, the material should not be allowed to erode.
(Section 3.c)
, Strengths:
- 1. The licensee's preparation and execution of the reactor vessel hydrostatic test was well organized and controlled. (Section 5.0)
- 2. The licensee's response to a January 17, 1988 reactor scram signal and subsequent equipment malfunctions was prompt, thorough and effective.
(Section 4.d)
- 3. Using non-nuclear steam for testing of high pressure coolant injection system and reactor core isolation cooling system enabled the licensee to discover problems which may not have been easily identifiable using nuclear steam due to radiological conditions. (Section 3.b)
l 4 i TABLE OF CONTENTS Page
- 1. Summary of Facility Activities ........................ 1 j
- 2. Followup on Previous Inspection Findings .............. 1 4
- 3. Routi ne Peri odi c Inspecti ons . . . . . . . . . . . . . . . . . . . . . . . . . .
Surveillance Testing
- a. l
- b. Radiation Protection and Chemistry
- c. Fire Protection ,
I l 15 4 Review of Plant Events ................................
- a. Spurious Isolations of RHR SFutdown Cooling System I
- b. Reactor Water Cleanup System Spurious Isolation '
- c. Engineered Safety Feature Actuations Due to a Failed Logic Relay
- d. Spurious Reactor Protection System Actuation
- 5. Review of Reactor Vessel Hydrostatic Test Procedure i 19 and Test Results........... ......................... i
- 6. Integrated Leak Rate Testing .......................... 21 Licensee Nuclear Organization Management j 7.
Realignment ......................................... 23 ! 24
- 8. Management Meetings . .................................
Attachment I - Persons Contacted Attachment II - Regional Administrator's Tour Observations
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DETAILS 1.0 Summary of Facility Activities The plant was shutdown on April 12, 1986 for unscheduled maintenance. On July 25,1986, Boston Edison announced that the outage would be extended to include refueling and completion of certain modifications. The reactor core was defueled on February 13, 1987. The licensee completed fuel re-load on October 14, 1987. Reinstallation of the reactor vessel internal - components and the vessel head was also subsequently completed. During this report period, the licensee performed the reactor vessel hydrostatic test and the primary containment integrated leak rate test On December 9, 1987, Pilgrim (ILRT) as described in Sections 5.0 and 6.0. Station conducted a partial participation emergency preparedness exercise. On December 14, 1987 the licensee announced as part of a planned manage-ment realignment, the appointment of eight managers to key management positions in the licensee nuclear organization at Pilgrim Station. The details of the management realignment are described in Section 7.0. NRC inspection activities during the report period included: 1) observa-tion of the licensee's annual emergency preparedness exercise on December 9, 1997, 2) NRC Reactor Operator Licensing examinations were administered to eight candidates on the week of December 7, 1987, 3) ob-servation of the primary containment ILRT and review of the test results during the week of December 21, 1987. The results of these inspections are documented in inspection reports 50-293/87-54, 50-293/87-56, and 50-293/87-5E. In addition, representatives of the NRC's Office of Inves-tigation were onsite December 3, December 7, and December 8, 1987 to interview onsite security personrol. On December 8, 1987, the NRC Regional Administrator for Region I, Mr. William T. Russell, toured the plant with the resident inspectors. On January 7, 1988, Dr. Thomas E. Murley, Director of the Office of Nuclear Reactor Regulation (NRR) and other NRC representatives toured the plant with the resident inspectors. 2.0 Followup on Previous Inspection Findings (Closed) Unresolved Item 82-24 Discrepancies in the Licensee's Response to IE Bulletin 79-08 Previous reviews of this item are documented in the inspection reports 50-293/82-30, 50-293/83-01, 50-293/83-14, and 50-293/84-26. IE Bulletin (IEB) 79-08 and the TMI Action Plan Item II.E.4.2 required licensees to review the containment isolation initiation design and procedures to ensure proper initiation of containment isolation, upon receipt of an automatic containment isolation signal. The licensee provided the ' results of their review in letters dated April 25, and August 21, 1979.
2 The licensee stated that the RBCCW supply and return lines, instrument air line, RHR to spent fuel pool cooling tie line, and torus make up line would be manually isolated and that station procedures would specify the requirements for manual isolation if a containment isolation signal was received. This was documented as acceptable by NRC:NRR in letters to the licensee dated December 18, 1979 and April 3, 1980. However, an inspector identified that manual i solation of these lines with qualified valves is not possible. Any valve which is used for primary containment isolation must meet Seismic Class I (FSAR section 12.2) and applicable Further, i f manual 10 CFR 50, Appendix J, containment leakage testing criteria. operation of a valve i s required to effect containment isolation, the isolation point for the valve must also be accessible under those condi-tions which make its use necessary. In response to the inspector's questions, the licensee re-evaluated their response to the IEB 79-08 and TMI Action Plan Item II.E.4.2, and concluded that isolation of these lines is assured by the use of Seismic Class I check valves. The licensee also agreed that isolation for the RBCCW supply line, instrument air line, RHR to spent fuel pool cooling tie line, and torus makeup line cannot be performed by manual valve closure. The RECCW return line from the drywell can meet the isolation valve criteria with MOV-4002 which is seismic class I, local leak rate tested and can be closed by a control switch located in the main control room. The licensee subsequently submitted a supplemen+.a1 response to IE Bulletin 79-08 and TMI Action Plan Item II.E.4.2 on October 24, 1984 correcting the previous response. The inspector reviewed the supplemental response and verified that the contents were consistent with the conclusions drawn from the j licensee's re-evaluation and the FSAR. Both RBCCW supply line and instru- I ment air line are considered Class C lines in Section 7.3 of the FSAR i since they penetrate containment but have no interaction with the primary containment free space or the reactor vessel. According to the original l design criteria, a single check valve is provided to attain isolation for a Class C line. These check valves are seismic class I and local leak rate tested. The inspector reviewed the results of local leak rate test data for these check valves which were performed on June 12 and July 26, 1987 and found no discrepancies. The torus makeup line is identified as Class B in Section 7.3 of the FSAR. The torus makeup line is non-essen-tial and ties the condensate transfer system into the RHR test line, which . I penetrate primary containment and ends below the torus water level. For water-sealed Class B lines such as sne torus makeup system, the original l plant design bases allow one isolation valve in addition to the water seal to meet isolation requirements. Also, the Safety Evaluation by the NRR on Appendix J Review indicate that Type C testing is not required for valves in lines which terminate below the level of the suppression pool. As for the RHR to spent fuel pool line, the licensee revised the operating pro-cedures 2.2.85, Fuel Pool Cooling and Filtering System, prohibiting the use of the RHR to spent fuel pool lines except i n cold shutdown. The inspector had no further questions. This item is closed. 4
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3 (Closed) Inspector Follow Item (IFI 87-27-02) - Cracking of Surge Ring Brackets in large GE Motors Cn July 2, 1987, IE Information Notice 87-30, Cracking of Surge Ring Brackets in large GE motors, was issued. The purpose of the notice was to alert recipients of a potential for f ailure of surge ring brackets and 7 cracking of felt blocks in large, vertical electric motors manufactured by General Electric Co. Felt blocks are used in large electric motors to keep the windings separated where they loop back at the end of the stator. The blocks are attached to a surge rirg that is held in place by L-shaped surge ring brackets welded to the surge ring and bolted to the motor cas-ing. Failure of these surge -ing brackets and cracking of the felt blocks allows movement and wear of the end-turns, leading to a reduction in insulation resistance and possible motor failure. In a0dition, broken pieces of the surge ring bracket ma3 enter the space between the stator and the rotor, resulting in electrical or mechanical motor degradation. i Following an investigation to determine the applicability of the subject notice to the Pilgrim Station, the licensee found that RHR, core spray, and recirculation pump motors were potentially affected. RHR and core ; spray pump motors were overhauled on site by GE u* der contract with the ! licensee in 1986. The surge ring brackets were net inspected during the cverhaul. However, small cracks were found on the "A" anc "C" RHR pump meter winding felt blocks. The amount of cracking found was disposittened by GE to be acceptable and a normal phenomenon found in form wound motors. On July 27 through August 5, 1987, GE performed a surge ring bracket inspection of the RHR and recirculation pump motors using a boroscope with I the motors in place. The inspection of the RHR motors (A thru D) revealed l absence of cracks on the surge ring brackets. During the inspection of
~
the "B" recirculation pump motor, it was noted that the recirc motor surge ring bracket construction is of the bolt and stud design, whereas the RHR and core spray motor brackets are of the L-shaped design. The L-shaped design configuration is known to have the potential of cracking, according to the IE Notice 87-30 and the GE letter to the licensee dated l July 14, 1987. l During the week of October 26, 1987, "B" core spray pump motor was dis-assembled and the surge ring brackets inspected by G.E. Due to the geo-metry of the core spray pump motor internals, there is limited access for ! the bore scope, therefore, this inspection could not be accomplished with-out partial disassembly of the motor. It was verified that the design had . 12 brackets per surge ring and two surge rings for the top end turn assem- } bly and two surge rings for the bottom end turn assembly. None of the ) brackets had indications of cracking. The licensee scheduled the inspec- j tion of the "A" core spray pump motor during the next outage because of i scheduling conflicts. The licensee indicated that based on the inspection l l l
)
P k 4 results of the RHR and "B" core spray pump motors, postponement of the "A" core spray pump motor inspection is justified. The licensee also added that the number of operating hours and starts are similar between the A and B core spray pump motors since both core spray systems' testing and surveillance requirements are similar. The inspector had no further questions. This item is closed. (Closed) Unresolved Item 87-45-05 Failure to Issue Licensee Event Reports In inspection report 50-293/87-45 the NRC identified three engineered safety feature actuations which appeared to be reportable under 10 CFR 50.73 but had not been reported by the licensee. The licensee reviewed the three actuations, agreed that they should have been reported and agreed to issue License Event Reports (LER) to document the occurrences. In addition the licensee agreed to perform a review of previous actua-tions to determine if any additional reports were needed. During this inspection period the licensee's compliance section conducted a review of all Failure and Malfunction Reports (F&MR) issued from April 1986 through the present. This review identified four F&MRs that fit the description of an ESF actuation under the current BECo interpretation of NUREG 1022. The licensee will submit LERs to document the following ESF actuations at a later date. 4/28/87 Initiation signal to both Emergency Diesel Generators (EOG) 6/7/87 Actuation of Reactor Building Isolation and Standby Gas Treatment System start signal 9/17/87 Auto start of "A" EDG f 10/6/87 Reactor Water Cleanup and Shutdown Cooling System Isolation These LERs will be reviewed upon issue as part of the normal resident inspection program. The inspector has reviewed the licensee's actions in addressing open item 87-45-05 and is satisfied that those actions were thorough and timely. This item is closed. 3.0 Routine periodic Inspections The inspectors routinely toured the facility during normal and backshift hours to assess general plant and equipment conditions, housekeeping, and adherence to fire pro:ection, security and radiological control measures. Inspectior.s were conducted between 10:00 p.m. and 6:00 a.m. on January 17, 18, and 19,1988 for a total of four hours and during the weekends of December 12, 19, 27, 1987 and January 3, 9, 17, 1988 for a total of 17 hours. Ongoing work activities were monitored to verify that they were i I 4 l
5 being conducted in accordance with approved administrative and technical procedures, and that proper communications with the control room staff had been established. The inspector observed valve, instrument and electrical equipment lineups in the field to ensure that they were consistent with system operability requirements and operating procedures. During tours of the control room the inspectors verified proper staffing, access control and operator attentiveness. Adherence to procedures and limiting conditions for operations was evaluated. The inspectors examined equipment lineup and operability, instrument traces and status of control room annunciators. Various control room logs and other available licensee documentation were reviewed. The inspector observed and reviewed outage, maintenance and problem inves-tigation activities to verify compliance with regulations, procedures, codes and standards. Involvement of QA/QC, safety tag use, personnel qualifications, fire protection precautions, retest requirements, and reportability were assessed. The inspector observed tests to verify performance in accordance with approved procedures and LCO's, collection of valid test results, removal and restoration of equipment, and deficiency review and resolution. Radiological controls were observed on a routine basis during the report-ing period. Standard industry radiological work practices, conformance to radiological control procedures and 10 CFR Part 20 requirements were observed. Independent surveys of radiological boundaries and random i surveys of nonradiological points throughout the f acility were taken by the inspector. Checks were made to determine whether security conditions met regulatory requirements, the physical security plan, and approved procedures. Those J checks included security staffing, protected and vital crea barriers, personnel identification, access control, badging, and compensatory ; measures when required. { l
- a. Surveillance Testing
-- Diesel Generator prelube pump Failure On December 13, 1987, the prelube pump for the "B" emergency diesel generator (EDG) failed to restart on demand during a routine surveillance test. Upon disassembly it was identified that a small piece of metal had become lodged between the pump rotor and idler gear. The interference from the metal caused the pump motor breaker to trip on pump start. An identical failure occurred during a lost of offsite power event on November 12, 1937. In that case the failure caused a lengthy delay in returning an idle diesel to service. While not required for diesel operation, the prelube system reduces EDG bearing wear during equipment start.
.~
6 In response to the failures, the licensee drained and inspected the lube oil sump, and disassembled and inspected the lube oil filters, strainers and heater. The lube oil heater was found to have failed in the energized mode resulting in significant carbon deposits in the heater and filter. No appreciable deposits were found in the lube oil sump. In addition, a piece of filter element packaging material was found in the lube oil filter housing. No foreign material which could have contrib-uted to the prelube pump f ailure, however, was found. The pump was replaced and the diesel was returned to service. No addi-tional failures occurred during the inspection period. The two - pumps which failed had in-sequence serial numbers. Licenses Quality Control personnel performed magnetic particle and dye-penetrant testing of the internals of a third in-sequence pump in the warehouse. No flaws were noted. The licensee is pursu-ing the root cause of the failures in cooperation with the pump vendor, Viking Pump. The licensee stated at the exit interview that the "A" EDG prelube pump and lube oil heater would Le inspected during the next "A" diesel outage. The inspector will continue to monitor licensee fellowup to this problem.
- Steam Testing of the High pressure Coolant Injection and Reactor Core Isolation Cooling Systems The licensee completed full pressure steam testing of the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) system turbines by utilizing temporary oil fired auxiliary boilers as a source of non-nuclear steam. The full pressure steam testing is part of r post-maintenance and system operability check. Both HPCI and RCIC systems were overhauled during the current outage. Utilizing temporary test procedures TP 87-198 and TP 87-199, the HPCI/RCIC testing included turbine overspeed trip, pump full flow capacity and operation from the alternate shutdown panels. Also during the test, the suction path was changed from the condensate storage tank to the torus and back.
During the testing, several problems were identified by the licensee in both HPCI and RCIC systems. In HPCI, problems with the governor control system were noted including a minor oil leak in the servo-motor. Steam leaks at gauges and turbine drain line were also discovered. In RCIC, the licensee dis-covered a previously installed blank flange in the turbine steam leak off line which caused steam leaks. A few problems were also noted on the RCIC governor control system. The licensee is in the process of dispositioning these items. The inspector noted that using non-nuclear steam for the testing enabled the licensee to discover problems which may not have been easily identifiable using nuclear steam due to the radiological condi-tiens. The inspector will review the results of the tests and dispositioning of the problems identified during the tests.
1 7
- Incorrect Installation of Fire Dampers -i On December 17, 1987, during performance of a routine surveil-lance test the licensee inadvertently actuated two fire dampers. ,
One of the dampers failed to fully close due to interference ' with a hook used to secure it in- the open position. When. the . fusible link was energized, the metal damper retaining . strap should have f allen away. allowing full closure The hook attach-ing the strap to the fusible link was oriented with the open : side toward the damper. The damper caught on'the hook and re- ' mained partially open. Upon discovery the licensee immediately stationed fire watches at all areas containing - suspect dampers. Inspections were promptly conducted and it was identified that ; The all of the installed hooks were oriented in this manner. ' hooks were repositioned so that the open side faces away from the damper. Three dampers were inaccessible and compensatory measures remain in place pending inspectic1. , The dampers were originally supplied to fle licensee without the hooks. A revision to the plant design change (PDC) package , adced the hooks to facilitate surveillance testing. Installa- < tion instructions contained in the PDC specified hook orienta- ,. tion with the open side toward the damper. The vendor data sheet supplied by Air Balance Inc. also showed the hook instal-led in this manner. Licensee event report (LER) 87-020-00 wts issued describing the j problem and corrective actions taken. The LER states that pre-liminary licensee assessment of the issue determined that it did ; ] not meet the reporting threshold of 10 CFR Part 21. The inspec-
- tor discussed the Part 21 reportability with the . licensee's Nuclear Engineering Department (NED). NED personnel stated that the f ailure mechanism was created by the licensee when the hook ;
was added. In addition the presence of mitigating factors such ; as fire detection and suppression, and control of combustible l materials support the conclusion that a substantial safety hazard did not exist. The licensee also feels that LER 87-020-00 contains sufficient information- to clearly define the problem. The inspector had no further questions in this area. The inspector examined two dampers in the cable spreading room to verify that the hooks had been reoriented. Both hooks had been modified, however, neither of the dampers had locking rings installed at the hook to retaining strap connection as required by the installation instructions in the PDC. The licensee-reviewed the function of the locking rings and concluded that they were not required. A change to the PDC was initiated to delete the ring. The inspector had no further questions. 1 4 4
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8
- b. Radiation Protection and Chemistry
- Locked High Radiation Area Access Control Ouring the period covered by inspection report 87-57, four ,
instances occurred in which the licensee failed to properly con-trol access to areas that had been designated as locked high radiation areas. In three of these cases, doors to locked high radiation areas were found closed but not locked and in the fourth case a door into a locked high radiation area was found to not be on the list of doors that were being controlled under the locked high radiation area door procedure. On December 15, 1987, a contract painter f ailed to check that the door to the locked high radiation area he was exiting was properly latched. The unlatched door was identified during tne nen routine check of high radiation area doors. Licensee per-sonnel immediately latched the door and initiated a radiological occurrence report (ROR) to document the occurrence and track all actions taken during the investigation. Surveys of the area showed no dose rates greater than 1000 miilirems per hour (MR/hr). Interviews with the individual involved determined that the procedures and requirements were well understood and that the HP technician had inferned them of their responsibil-ities prior to entry into the area. On December 27, 1987, and again on January 8,1988, instances similar to the one described above took place. In both cases the licensee initiated RORs and took steps to determine: 1) who t had been in the area, 2) were they aware of the procedure, and l
- 3) had they been properly briefed prior to entry into the areas involved. In both of these cases the root cause has been deter-mined as personnel error.
In one instance the licensee identified that one of the multiple doors into an area classified as a locked high radiation area was not on the list of doors to be checked on a routine basis. The door was immediately checked and found to be locked. Records have been audited to determine if any unauthorized entry into the area had occurred and no instances were identified. The door has been placed on the list and is now routinely checked. The inspector reviewed licensee actions as a result of these instances and is satisfied that in all cases, the immediate and followup actions were timely and complete. Surveys taken were comprehensive and conducted almost immediately after discovery of unlocked areas. Oose calculations were performed and dosimetry read in all cases. Involvement by senior HP and plant management was evident in all instances.
9 Inadequate control of locked high radiation areas has been an , area of longstanding NRC concern. Notices of Violation have been issued in the past, during inspections 50-293/87-03, 50-293/87-11, and 50-293/87-19 which addressed these concerns. In regard to these violations the licensee instituted corrective actions which have been successful in addressing segments of the , problem but have not been successful in preventing recurrence of events involving high radiation area door control. The inspector has independently reviewed the licensee's program for control of high radiation areas and high radiation area key control and has found them adequate. Although the programs themselves are adequate and personnel have been trained on those programs, instances still occur where locked high radiation areas are not adequately controlled. l Based on review of these four instances coupled with the review of Unresolved Item 87-50-08, the inspector deterrnined that the licensee actions in response to these previous findings have not prevented recurrence. Failure to comply with the requirements i of Technical Specification 6.11 and Implementing Procedure 6.1-012 is an apparent violation of NRC requirements as docu- , mented in Appendix A of the cover letter to this report (87-57-01). Licensee response to Appendix A should include those measures taken to insure that corrective actions are effective and lasting.
- Contaminated Clothing Offsite On December 17, 1987, at 7:26 p.m. hours a Bechtel pipefitter who was exiting the reactor building, set off a whole body por-i tal monitor alarm. The portal monitor indicated contamination of his chest area and lef t hand. The health physics technician ,
on duty at the access point removed the individual from the por- l tal monitor and began performing a survey using a RM-14 with DT ' 260 probe. The HP technician identified; 1) contamination on i the individual's left hand, 1-2 thousand dpm per 100 square l
- centimeters (K DPM), which was removed by washing, 2) contamina- l 1
tion on the shirt in both the chest (80K DPM) and lower stomach area (1K DPM). The shirt contamination was removed by tape (80K The employee, OPM) and washing with soap and water (1K DPM). now wearing an undershirt and trousers, was then sent to clear the portal monitor which again alarmed and indicated contamina- - tion in the chest area. The HP technician again surveyed the individual and identified contamination on the undershirt in the chest area (70K DPM). The individual was then sent into the portal monitor bare chested and was cleared. The individual was given his outer shirt, which was still wet from decontamination , and cleared through portal monitor. At this point, the indi-vidual removed the wet shirt, put on his jacket, cleared the portal monitor again, and left for his home. ?
10 Upon returning to work December 18, 1987, the individual was given a whole body count to determine if any internal contamin-ation had occurred. The whole body count showed no internal contamination. After completion of the whole body count the individual was interviewed to determine how he had been contam-inated, where the occurrence took place and how long he was contaminated prior to detection, to calculate skin dose received. The interview revealed that the individual had been contaminated when he disconnected a partially pressurized service air hose and depressurized it. The interview also revealed that the individual used the portal monitor at the 91 ft. elevation of the reactor building, received an alarm, did not call for HP assistance but instead tried to decontaminate himself prior to proceeding to the reactor building access. Station procedures require that an individual who finds himself contaminated is to call health physics for assistance. The individual stated that he was aware of this requirement. During the interview the inchidual expressed concern about whether his heavy winter jacket could have shielded the contamination on his shirt and uncershirt f rom detection by the cortal monitors. To demon-strate that this could not happen, a HD supervisor placed plastic bags, which contained the contamination removed from his shirt, inside the coat and attempted te exit through twe Fr-tals. The portal monitors alarmed on each attempt. The int f-vicual appeared satisfied with the demonstration put his jacket back on, with the plastic bags removed and attempted to leave the reactor building. An alarm was actuated on the portal conitor and contamination was indicated on the left arm. The en duty HP technician removed the individual f rom the portal monitor and identified 3K DPM contamination on the upper right - sleeve (outside) of the jacket even though the jacket had not been worn into the reactor building. At this juncture the indi-vidual expressed concern over whether the shirt that he had worn the previous day could still be contaminated. The licensee had a HP technician accompany the individual to his home. The individual's shirt was found to be contaminated, was bagged and returned to the site. Surveys of the individual's home and vehicle identified no further contamination. J Efforts to determine how the contaminated shirt was worn through the portal monitors without setting of an tiarm yielded positive results. The individual stated that he had purposely kept him-self away from the portal monitor in an attempt to keep his wet shirt away f rom his skin. The licensee taped the plastic bags, with the contamination in them, back onto the shirt and an HP supervisor attempted to pass through the portal monitors by l l 1
I i I l 11 j i mimicking the body posture used by the individual when he cleared the monitor. The HP supervisor was able to pass through six dif ferent monitors without setting of f an alarm. The HP super-visor then used the portal monitors in the correct manner and all six monitors alarmed proving that the equipment was func-tional. The licensee has evaluated the occurrence to identify the root causes and immediately implemented corrective action. This occurrence was caused by one sequence of events that involved two distinct personnel errors. The primary cause involved the . failure of the HP technician to perform an adequate survey of the contaminated individual's clothing when the portal monitor alarm was received. The second problem involved the failure to properly use the installed portal monitors at the reactor build-ing access. In addition to personnel interviews to identify the sequence of events the licensee also reviewed procedural adequacy, personnel training and portal monitor calibration and performance. These reviews verified ttat training was adequate and portal monitor performance was as designed. Procedures for control of contam-inated individuals at the reactor building access did not spec-ifically require that all articles of clothing require a 100% frisk prior to this occurrence. Instructions have been posted at the reactor building access which now clarify the procedure to be followed when an individual is found to be contaminated. The portal monitors in use at Pilgrim do not presently have a switch at chest level which must be actuated to start the moni-toring process. Lack of this feature alicaed the individual wearing a contaminated shirt to lean away from the machine suf-ficiently to clear the monitor without any alarm. The licensee has determined that the manuf acture of the portal monitor now produces a chest high switch for the installed model and will install them in the future. Calculations have been performed by the licensee to determine the radiation dose received by the individual and the amount of radioactive material that was released from the site on the con-taminated shirt. The results of these calculations show that the individual received a localized radiation dose to the skin of 260 Mrem, which is below the federal limits for skin exposure, and that the amount of radioactive material on the individuals clothing was 0.2 microcuries which meets the federal criteria as an exempt quantity of Co-60. The inspector is satisfied with the licensee's analysis and corrective actions and has no further questions.
12
- Allegation of Improper Disposal of Radioactively Contaminated Shrubs ( RI-87-A-0107)
On August 31 and September 11, 1987, the NRC resident office at Pilgrim received allegations that radioactively contaminated shrubs had been removed from the site and improperly disposed. The alleged improper disposal occurred on July 23, August 26 and August 28, 1937. During this time period the licensee removed a large number of shrubs from various areas of the site, including those planted near the old administration building and the switchyard. The shrubs were removed to facilitate site con-struction activities and to alleviate certain security concerns. Upon receipt of the first allegation on August 31, 1987 the NRC requested that the licensee perform an evaluation and provide the results for review. In addition an independent NRC review was initiated. Resident and specialist inspectors reviewed the licensee's con-clusions. The licensee evaluated material release records and interviewed personnel regarding removal of shrubs during the week of July 20, 1937. Several truckloads of shrubs that were transported of f site during the midnight shift on July 24 were examined in detail . Because trace amounts of Cobalt-60 had pre-viously been found in soil onsite, some of the shrubs had the soil removed from the roots prior to release. Each shrub was hand surveyed and found to meet established offsite release criteria. They were transported first to the licensee's shore-front area and later to a dump site on licensee property. The licensee concluded that the shrubs had been adequately surveyed and that no radioactive material had been improperly released. The resident inspectors reviewed the licensee's program for con-trol of release of material from the site. This area was also evaluated by NRC specialist inspectors during inspection 50-293/ 87-19. Both inspections concluded that appropriate surveys and release limits have been established and implemented. Resident and specialist inspectors examined licensee release records for the dates in question to verify that vehicles leaving the pro-tected area had been properly surveyed. No discrepancies were identified. An NRC resident inspector accompanied by a licensee representative collected four samples of the shrubs which had been deposited in the dump site discussed above. Each of the four samples consisted of root, branch and foliage clippings from a number of different shrubs. The samples were indepen-dently analyzed by the NRC, Three of the samples indicated no contamination. One sample indicated only trace levels of Cobalt
-60. Measurements showed that the amount of CO-60 present in this sample was about 2% of the average radioactivity typically found in soil due to naturally occurring isotopes.
13 The licensee's program for release of material from the site appears adequate. Appropriate survey techniques and release limits have been established. Review of records confirmed that the program is being implemented. Samples of the shrubs col-lected by the NRC showed zero or negligible contamination and pose no health and safety concern. Based on the above this allegation is considered closed. NRC Region I staff provided status briefings concerning this allegation to Senator Kennedy's staff and to the Massachusetts Department of Public Health.
- Allegation of Airborne Radioactivity in the Trash Compaction Faci 11 :y (RI-87-A-0120)
On October 5, 1987, the resident office received an anonymous allegation that personnel working at the sort table in the trash compaction facility (TCF) were being routinely exposed to air-borne radioactive contamination. The alleger stated that the two filter systems designed to treat exhaust air from the sort table prior to dischtrge into the room were not functioning, and that the filter differential pressure alarm circuits had been disabled. On October 7 and 8,1987, NRC specialist inspectors toured the TCF and examined the design and condition of the equipment. The sort table is used to separate contaminated materials for com-pattion and disposal. Potentially contaminated air is exhausted from the table, passed through two filters operating in parallel, and released into the room. Airborne radiation levels in the room are measured by means of a separate air monitor which is operated whenever the sorting table is used. The alarm is typically set at 3 X 10 -10 (3E-10) microcuries per cubic cen-timeter (cc). In addition the filters are surveyed daily and changed if contact dose rates exceed 2mR per hour. The inspec- : tors also examined the trash compaction unit in the area and found that similar controls had been applied, Based on the above, no immediate health and safety concerns were indicated. ] On January 15, 1988, the resident inspectors toured the TCF, ) examined equipment operation and interviewed licensee and con- j tractor personnel involved in ongoing work activities. A radis-tion work permit specifying protective clothing, health physics coverage, and use of a continuous air monitor is in place to control work at the sort tabic. Personnel involved stated that trash bags were surveyed prior to sorting and rejected if radia-tion levels exceeded Smr/hr, if they contained liquid, or if any powdery material was present. The health physics technician on duty stated that filter radiation levels are monitored daily. l
14 J Workers and health physics personnel also stated that filter dif ferential pressure (dp) instruments are monitored to detect filter plugging, however no one had been clearly assigned this responsibility and no dp limit was established. The inspector observed the operation of the table and noted that the "filter restricted" alarm actuated for one of the two filters. The alarm actuated for the filter displaying the lower dif ferer.tial pressure. When questioned workers statec that much of the monitoring and alarm circuitry for the table was not functional, and that the filter alarm was not reliable. The table was originally part of a larger processing system and much of the disconnected circuitry was intended to perform functions which are no longer needed. The inspector verified that current filter dp readings are consistent with the manufactures name plate data. It appears that the general process applied, including inspec-tion and survey of trash bags prior to sorting, daily filter surveys and continuous air monitoring would preclude airborne radioactivity problems. Based on the above this allegation is closed. However, the inspector noted that no work instructions existed describing the controls applied and equipment monitoring requirements. When discussed with licensee radiation protection management they promptly committed to review the situation and issue appropriate guidance. This was confirmed during the inspector's exit interview. Erosion of Construction Dirt into Wetland On January 15, 1988, at 5:45 p.m. the licensee made an ENS notification in accordance with 10 CFR 50.77 (b)(2)(vi) which requires the licensee to inform the NRC of an event or situation related to health and safety of public for which a news release was made or notification of another government agency has been made. During routine environmental monitoring, the licensee observed erosion from a pile of construction dirt into an adja-cent licensee controlled wetland. The Plymouth Conservation l Commission and the Massachusetts Department of Public Health were notified and the press release was made by the licensee. Also on January 16, 1988 two representatives from the Plymouth Conservation Commission toured the area. l In the last several years during onsite excavation for plant ! modifications, dirt, asphalt and concrete containing low levels of contamination were stored in a fenced in storage area outside the protected area on the licensee's property. The licensee estimated that the storage area contains 110,000 cubic feet of I material. Before removal f rom the protected area, samples of { l l i
15 material were obtained and isotopic analyses was performed by the licensee. The activity found was reasonably uniform at levels of 10(1E-6) and 10(IE-7) microcuries of Cobalt-60 and Cesium-137 per gram. Sampl1ng and storage of this material was previously reviened during inspection 50-293/87-18. On January 21, 1983 the inspector toured the area, accompanied by a licensee health physics technician, and performed a survey of the storage area and found no detectable radiation above back-ground levels. During the tour the inspector noted that bales of hay had been put around the perimeter of the fence which borders wetlands area to prevent further erosion of material. The fenced in storage area was secured with a locked gate. The inspector's survey of the area and review of licensee's analyses indicate that the level of activity does not represent a health or safety concern. However, the inspector raised a concern to the licensee management that the material should not be allowed to erode. The inspectors will continue to monitor the licensee actions in formulating long term solution to properly dispose of the material.
- c. Fire prctection On January 17, 1933, at 4: 55 a.m. the control room received a report from a security guard of smoke coming from a contractor lavatory trailer, which is located adjacent to the Bechtel warehouse inside the protected area fence. The onshif t fire brigade chief was dis- ,
patched to the scene and confirmed smoke and smoldering in the area. The fire brigade was immediately dispatched and fire was extinguished using a portable dry chemical extinguisher and a hose from a nearby hydrant house. Electrical maintenance was called to shut off the power to the trailer. By 5:30 a.m. , the fire brigade members had cleared the scene and a continuous fire watch was posted in the area. The cause of the fire was believed to be overheating of an overhead heating unit for the trailer, ho personnel injury occurred. The , inspector toured the scene with a licensee fire protection engineer on January 18, 1988. Minor damage to a small area of the ceiling in the trailer was observed. The Plymouth Fire Department was notified , by the licensee in the morning of January 18, 1988. l l 4.0 Review of Plant Events The inspectors followed up on events occurring during the period to deter-mine if licensee response was thorough and effective. Independent reviews of the events were conducted to verify the accuracy and completeness of licensee information. 1 j
16
- a. Spurious Isolations of RHR Shutdown Cooling System On December 7,1987, at 2:28 p.m. , an inadvertent isolation of both inboard and outboard containment isolation valves on the RHR shutdowi cooling suction line occurred. Prepsration for the reactor vessei hydrostatic test was in progress. As part of the hydrostatic test procGdure, a technician was installing an electrical jumper in the primary containment isolation system logic panel C-941 to bypass the '
reactor coolant system (RCS) high pressure interlock on the inboard isolation valve. When the termination screws were loosened to in- ' stall the jumper, the leads lost contact and caused a false high pressure isolation signal. RHR was in its shutdown cooling mode when the isolation signal was generated, and the shutdown cooling suction valves (MOV 1001-47, 1000-50) automatically closed as designed. Coincident with the closure of the valves, the "A" and "C" RHR pumps tripped automatically to protect the pumps from loss of adequate suction. The licensee determined the actuation was due to a person-nel error. The licensee revised Procedure 2.1.8.1, Class I System Hydrostatic Test, to caution the I&C technician of potential isola- i tion of RHR shutdown cooling system while installing the jumper. On December 8, 1987, at 9:45 p.m., the inbo.rd isolation valve (MOV 1001-50) on the RHR shutdown cooling suction line automatically l The automatic isolation occurred when the plant reached ; closed. 100 psig during pressurization for performance of the class I hydro-static test. The outboard isolation valve (MOV 1001-47) was already closed to form a pressure boundary for the test. The licensee's investigation determined that the cause of the isolation was that 4 Procedure 2.1.8.1 did not identify all the jumpers necessary to bypass the RCS high pressure interlock on the inboard isolation valve. As immediate corrective action, the licensee halted the pressuriza-i tion of RCS and reviewed the logic prints. The licensee revised l 4 Procedure 2.1.8.1 to reflect the need to install an additional jumper 1 in panel C-942. In reviewing this event along with other similar l events documented in previous inspection reports, the inspector noted i that inadequate planning and inadequate procedures appear to be a I common root cause for several ESF actuations which occurred on September 17, September 22, October 15 and October 24, 1937. The . inspector expressed this concern at the exit meeting with licensee management. The licensee informed the inspector that the Technical Group is in the process of developing generic guidance for isolating or jumpering an electrical component which may cause inadvertent safety system actuations. The inspector will continue to monitor the effectiveness of licensee's corrective 3ction to prevent further ESF actuations due to inadequate planning and inadequate procedures. l l i i
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- b. Reactor Water Cleanuo System Spurious Isolation i
On December 17, 1987, at 11:05 a.m., the inboard primary containment isolation valve on the reactor water cleanup (RWCU) system suction l line automatically isolated. I&C technicians conducting a routine surveillance of the RWCU high- area temperature isolation logic inad-vertently grounded a lead which had been lif ted during the test. Grounding the lead resulted in a blown logic power fuse and isolation of the valve (MOV 1201-2). Following investigation by the control room supervisor, the fuse was replaced and the isolation was reset. The licensee's investigation concluded that the root cause is a per-sonnel error. The licensee informed the inspector that the proced-ure, 8 M.2-1.2.2, Reactor Water Cleanup Area High Temperature, will be revised to provide cautions to the control room operators and the I&C technicians. Also, an effort is ongoing to review recent ESF appropriate actuations caused by personnel error to formulate corrective actions,
- c. En;ineered Safety Feature Actuations Due to a Failed Logic Relay On January 6, 1988, at 2:50 p.m. , the coil of primary containment isolation system (PCIS) electrical relay 16A-K57 failed, creating a fault anJ resulting in blown logic power fuses. The deenergization of this portion of the PCIS logic caused a partial primary contain-ment isolation along with a reactor building isolation and start of the "B" Standby Gas Treatment system (SBGT). The licensee notified the NRC at 5:12 p.m. via ENS. The failed relay was a GE type CR120A relay. The licensee has experienced several failures of this type of i relay in the last few years. The licensee's evaluation of this high f ailure rate and corrective actions to address it are described in the inspection report 50-293/87-50.
On January 7,1988, the inspector reviewed maintenance request (MR) 88-9 which had been initiated to investigate the c<use of the above j mentioned ESF actuations and to replace the blown fuse and the faulty relay. The inspector noted that the relay replacement was performed using only procedure 3.M.1-11, Routine Maintenance. This procedure l contains general guidance and its stated use is for performing main-tenance activities which are not complicated or :riticsl enough to l require detailed written procedures. In this case, no step-by-step instruction was initiated to control the sequence of work, to control and tag lifted leads and jumpers, anc to enr.are verification and independent verification of system restoration. A similar problem ! involving lack of a suf ficiently detailed controlling procedure and the appropriate reviews during an electric relay replacement on
' November 24, 1987 was the subject of a violation as documented in the i
inspection report 50-293/87-50. The licensee informed the inspector that tne corrective actions to address the violation are being 4 formulated and will be submitted to the NRC. I l
18
- d. ,5,p;gr,ious Reactor Protection System Actuation On January 17, 1988, at 1:13 a.m., a spurious reactor scram signal was generated during the performance of a reactor level instrument calibration. The full scram signal on low water level was received due to a disturbence in the reactor water instrument line when an I&C technician was valving a level instrument (LI-263-59A) back in ser-vice. The Rosemount level transmitters (LT-263-57 A&B) which initi-ated the scram signal are on the same instrument rack. The licen-see's preliminary investigation indicated that the root cause of the event is attributed to a combination of personnel error and inade-quate procedure. The investigation also identified that the level instruments (LI-263-59 A&B) were incorrectly installed in that the sensing lines were reversed. The new Barton level instruments (LI-263-59 A&B) were recently installed during this outage and would only be used for local indication during a shutdown from outside the control room, The licensee is currently reviewing the plant design change (PD 85-07) records and post-installation test data to deter-mine the cause. Surveillance test records are also being reviewed by the licensee, This item is unresolved pending the completion of the licensee investigation (87-57-02).
Upon receiving the spurious scram the control room staff noted that scram discharge instrument volume (SDIV) vent valve CV302-23B primary containment vent and purge valves A05044B and A05035B and one of two redundant secondary containment isolation dampers in each line did not close. In addition the "B" standby gas treatment system (SGTS) did not start. Bued on the initiating event, these components should have actuated. The licensee notified the NRC of the failures via ENS at 5:00 a.m. on January 17, 1988. The control room staff conducted an immediate critique with available ItC personnel, and documented observations for management followup. Later on January 17, the licensee inspected the physical condition of the SDIV vent and drain valves and noted paint on the stem of CV302-23B, The paint was removed and the valve successfully stroke timed. The licensee held a second critique with management repre-sentatives on the morning of January 18, 1988 te assess the situa-tien. Subsequently, a walkdovin of involved isolation logie components was performed to verify relay contact configuration and to identify l any jumpers or lif ted leads. This walkdown was performed to the i extent possible without disturbing components. No discrepancies were l noted. Early on January 19, the licensee performed a test in which a I reactor scram was intentionally initiated. The same equipment failed i to actuato as during the January 17 scram. Pted on this licensee management stopped all work on the affected ca r!Nents. A task force i composed of members from the technical staff. Etems group, I&C and operations was designated to investigate ;',9 incident. This team reviewed available information and developed 4,n action plan. I I
19 Walkdowns of the air system piping and components supplying motive air to -SDIV vent valve CV302-23B were performed to verify that the as built configuration is in accordance with design documents and that components are in good physical condition. No discrepancies were identified. Valves CV302-23B and CV302-22B are supplied air by the same solenoid operated valves. The licensee deenergized these solenoid valves and observed that CV302-22B closed while CV302-23B did not. This indicates a mechanical problem with the valve or operator. The licensee was identifying replacement parts and pre-paring to disassemble the valve by the close of the inspection period. The inspectors will continue to monitor licensec followup to this failure. Licensee review of logic drawings confirmed that the remaining equip-ment which had not properly actuated shared common isolation logic components. A series of surveillance tests was performed to allow monitoring of key relay actuations. A single contact on a General Electric (GE) HFA relay was determined to be misfunctioning. The contact is required to close when an isolation signal is received, actuating the affected equipment. However, contact resistance remained high with the contact closed. The relay was replaced and the system successfully tested. The licensee contacted GE to coor-dinate disassembly and inspection of the relay. Dissassembly had not begun by the close of the inspection period. The inspector will continue to monitor licensee investigation of this failure. The inspector expressed concern that three separate equipment mal-functions had occurred during the inadvertent actuation. This may reflect weakness in the surveillance and post-work test program. However, the licensee's response to the actuation and subsequent malfunctions was prompt, thorough and effective. Control room oper-ators quickly recognized each of the failures. They held a critique on the same shif t with involved personnel. Critique observations . were clearly tocumented and provided to management. An additional critique with management present established priorities. Action was taken to freeze (quipment until an investigation plan could be developed and implemented. Followup was well coordinated and in-volved representatives of several portions of the organization. In
- this case licensee commitment to determining and correcting the i problem root cause was evident.
5.0 Review of Reactor Vessel Hydrostatic Test Procedure and Test Results During the inspection period the licensee completed the reactor vessel hydrostatic test. Several reactor vessel instrument nozzles were repaired during this outage, prompting performance of a hydrostatic test rather than a system leakage test. The reactor vessel reached minimum test pressure and all inspections were completed on December 9,1937. Only J ' minor leakage associated with mechanical connections, such as flanges and valve packing was identified. The reactor vessel was depressurized on December 12, 1987 af ter completion of excess flow check valve testing. . l I i l l 1
,_,J
20 The inspector reviewed the licensee's hydrostatic test procadure to verify that appropriate prerequisites, precautions and instructions had been included. A sample of valve lineups was reviewed to determine the ade-quacy of established test boundaries. Completed valve lineups were also examined. Control of temporary elect:ical and mechanical jumpers was evaluated to ensure proper documentation and restoration. The inspector observed installed pressure instrumentation and verified appropriate range and calibration status. The adequacy of staf fing to support test per-formance was periodically vertfied. The inspector reviewed test results and discussed tnem with engineering, operations, and quality control personnel to ensure that test changes were properly processed, adequate inspections were conducted, and that inspection results were promptly dispositioned. The licensee's preparation for and execution of the test was generally well organized and controlled. Procedures for test performance and con-duct of visual inspections were clear and comprehensive. A detailed Quality Control (CC) work instruction was developed specifying components and piping requiring inspection. Inspection assignments were broken down by location, elevation and component. This QC instruction also included a series of piping diagrams depicting the test boundaries which were ut.ilized to assist in inspection performance and documentation. The licensee's Technical Engineering Section, Quality Control staff and Nuclear Engineering Department each reviewed test boundary adequacy. In-spection , ssults were well documented, and maintenance requests were promptly iritiated to correct identified leakage. The licensee experienced two shutdown cooling isolations during implemen-tation of tne test procedure. These isolations are discussed in detail in section 4.a of this report. During the test the licensee identified leak-age past the seal ring at the stuffing box to pump casing joint on both recirculation pumps. Leakage flow was estimated to be one to two gallons per minute for each pump, The leakage wet the pump casings and portions of the suction piping, and acceptable inspections could not be completed in these areas. The licensee stated that similar leakage on at least one of the pumps was noted during the last outage. That leak sealed as system temperature increased during startup. The licensee believes that the leakage observed during the recent test will also stop as temperature is increased, and no pump repairs are planned. The licensee stated at the inspector's exit interview that the pump casings and suction piping will be reinspected during startup. The inspector noted that the test procedure did not contain valve lineups I for manual instrument isolation valves within the test boundary. Many i instruments and a significant portion of instrument piping has been replaced this outage. Visual inspections were performed of class I piping j downstream of these valves. The inspector questioned the basis for licen- j see confidence in instrument line isolation valve positions during the ' test. The licensee pointed out that hydrostatic testing of these lines i was not required during this outage. In addition excess flow check valve l l l
- _. = -, . . . - . _ - . . -
l 21 i testing was conducted immediately after completion of the hydrostatic test with the system still pressurized. Successful completion of the check valve testing requires proper alignment of the manual isolation valves, and provides assurance that the piping was pressurized during the visual inspections. The licensee however, agreed that the intent of the test had . been te pressurize and inspect this piping and that the current procedure does not adequately assure the correct valve alignment. Licensee managa-ment stated that the procedure would be revised to address this weakness. 6.0 Integrated Leak Rate Testing , On December 21, 1987, the licensee began performance of the primary con-tainment integrated leak rate test (ILRT). The containment was pressur-1:ed with air to the full test pressure of 45 pounds per square inch and maintained at this pressure for 24 nours. The 24 hour test period started at 10:15 p.m. on December 21, 1957. A regional specialist inspector was onsite during the ILRT to review the adequacy of the test procedure and to observe the conduct of the test. The preliminary licensee test results indicated a successful test, with measured leakage slightly greater than 20 percent of the allowable leakage. A primary contributor to the ob-served leakage was identified as a drywell pressure transmitter piping cap which had not been fully tightened. Upon completion of the specialist
.! inspector's review of the ILRT results, inspection report 50-293/87-58 will be issued documenting the inspectors findings.
While preparing for the primary containment integrated leak rate test (ILRT) the licensee observed that several torus temperature and moisture elements were not functioning properly. Troubleshooting identified cir-cuit faults at a torus electrical penetration assembly. The licensee ! removed the penetration assembly protective cover inside the torus and found that it was filled with water. The penetration is installed ver-tically through the top of the torus. On both the inboard and outboard sides of the penetration a metal frame is attached on which 28 terminal boards are mounted. Cables passing through the penetration, and supplying instrumentation in the torus also landed on these terminal boards. A protective cover is bolted over the frame and terminal boards on both i sides of the penetration. Design drawings specify that cover joints are l to be sealed with silicone tape. The licensee stated that the protective cover had not been properly sealed, allowing water intrusion and buildup. The water caused significant corrosion of the cable connectors, terminal boards and metal framework. This corrosion and water buildup resulted in l the observed electrical circuit faults. Licensee inspection of the other j torus electrical penetration identified similar conditions. Temporary i repairs of the temperature and moisture elements were made to allow ILRT i performance. Cables for communications, lighting, and torus to drywell l vacuum breaker indication also run through the penetration. The penetra-tion is not considered by the licensee to require environmental qualifica- l tion but is designated as a "Q" component. The licensee is evaluating the ) root cause of the water intrusion and is developing a temporary procedure ! to control repair and testing of the penetration. The inspectors will J continue to monitor licensee followup and corrective actions. l i l .I i
l 22 1 The licensee informed the inspector that penetration repairs would not be completed until after ILRT performance. The inspector questioned the ef fect of the planned repairs on the penetration leak tightness, and the ability to perform adequate leakage test af ter the planned rework. The licensee stated that the work would not affect penetration leakage but that adequate testing could be performed after work completion. Based on available drawings however, the licensee could not demonstrate adequate tastability. In response to NRC concern the licensee obtained the needed drawings from the vendor and verified that the penetration was completely testable. The inspector had no further questions. During the ILRT, the licensee identified a water leak in the high pressure coolant injection (HPCI) turbine room. It was determined that the in-creasing pressure in the torus air space caused the suppression pool water ! to back up through the HPCI turbine exhaust line and through the drain piping, overflowing the HPCI gland seal condenser onto the HPCI room floor. The turbine exhaust line discharges to the torus through a check valve and a locked open stop-check valve. To prevent any condensation 4 from collecting in the turbine exhaust line downstream of the check valve, a drain piping drains any condensation to the HPCI gland seal condenser through a drain pot. Two solenoid operated drain valves on the drain pot close automatically on a HPCI (Group IV) isolation signal. This is to orovide the isolation from the torus to the gland seal condenser. The licensee's investigation determined that leads had been lifted in the HPCI isolation interlock logic circuit since October 30, 1987 in support of the HPCI steam testing utilizing temporary oil-fired auxiliary boilers. With
. the HPCI isolation signal bypassed, the drain valves remained open as the drain pot was filled with the suppression pool water. The licensee sub-secuently relanded the leads in the HPCI isolation interlock logic circuit and the drain valves closed.
After reviewing the ILRT procedure, HPCI test procedure and interviewing licensee personnel, the inspector concluded that licensee review of the active maintenance requests prior to the ILRT was not thorough in that the lifted leads controlled by the MR 87-663 were not identified. The MR tags were attached on the HPCI isolation logic circuit inside a logic panel and thus the tags were not identified during a system walkdown prior to the ' ILRT. The drain valve positions were verified by the light indications on the control room panel 903 as prescribed in the ILRT procedure. The inspector also determined that the maintenance request above may not be an adequate method of identifying and tracking jumpers and lifted leads, especially for a long term application and for components which could affect other ongoing maintenance or surveillance. Station proce-dures do not require temporary modification controls for jumpers and lifted leads which are controlled by active maintenance requests. The inspector discussed these findings at the exit interview with licensee management. The licensee informed the inspector that a lifted leads and i ju per log will be kept in the control room to aid the operators in con-
- trolling lifted leads and jumpers.
4
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23 7.0 Licensee Nuclear Organization Management Realignment On December 14, and on December 31, 1937, the Boston Edison Co. announced, as part of a planned realignment occurring over the next several weeks, the appointment of the following managers to key management positions in the licensee nuclear organization at Pilgrim Station. Mr. Kenneth L. Highfill was named to assume the new position of Station Director. In this capacity, Mr. Highfill will oversee day to day operation of the Pilgrim Station including plant operations, planning and outage, nuclear training, plant support functions, and administrative services. Mr. Highfill will report directly to Mr. Ralph G. Bird, Senior Vice President-Nuclear.
-- Mr. Robert J. Barrett was named the new Plant Manager. Mr. Barrett will report to Mr. Highfill, the Station Director.
-- Mr. Roy Anderson, currently Deputy Outage Manager, was named to assume the new position of Planning and Outage Manager. Mr. Anderson will report to Mr. Highfill, the Station Director.
-- Mr. Ed Kraf t was named to assume the new position of Plant Support Manager. In this capacity, Mr. Kraft will oversee radiological, security, industrial safety and fire protection, and other station support functions. Mr. Kraft will report to Mr. Highfill, the Station Director. ,
-- Mr. Donald Gillespie, currently Director of Planning and Restart, was appointed to the position of Quality Assurance Departeent Manager. '
Mr. Gillespie will assume the position after completing his Senior ] Reactor Operator training. The Quality Assurance Department Manager ) reports to Mr. J. E. Howard, Vice President-Engineering. Mr. Frank Famulari, currently Operations Quality Control Group Leader, was named to assume the newly created position of Deputy ; Quality Assurance Department Manager. Mr. Famulari will report to Mr. Gillespie, and be acting Department Manager until Mr. Gillespie assumes the position after completing the Senior Reactor Operator
; training.
-- Mr. John F. Alexander was named to assume the position of Operations
. Section Manager. Mr. Alexander will report to Mr. Barrett, the Plant j Manager. i
-- Mr. Donald J. Long was named Security Section Manager. Mr. Long will ~
report to Mr. Kraft, the Plant Support Manager, i I I 3 1
24 8.0 Management Meetings At periodic intervals during the course of the inspection period, meetings were held with senior facility management to discuss the inspection scope and preliminary findings of the resident inspectors. On January 26, 1988, the inspectors conducted a final inspection exit interview to formally present inspection findings.
Attachment I to Inspection Report 50-293/87-57 Persons Contacted
- R. Bird, Senior Vice President - Nuclear
- K. Highfill, Station Director K. Roberts, Plant Manager R. Barrett, Deputy Plant Manager R. Anderson, Planning and Outage Manager E. Kraft, Plant Support Manager F. Famulari, Deputy Quality Assurance Manager D. Swanson, Nuclear Engineering Department Manager J. Alexander, Operations Manager N. Brosee, Maintenance Manager J. Jens, Radiological Protection Manager J. Seery, Technical Manager R. Gra:io, Field Engineering Manager P. Mastrangelo, Chief Operating Engineer R. Sherry, Chief Maintenance Engineer N. Gannen, Chief Radiological Engineer D. Long, Security Manager F. We:niak, Fire Protection Manager
' Senior licensee representatives present at the exit meeting.
_ . ._ _ . _ _ _ _ _. ._ _- - . ~ _ _. 4 I i i t ATTACHMENT II i -l January 6, 1988 ' s MEMORANDUM FOR: Ken Roberts , Plant Manager FROM: Clay Warren Senior Resident Inspector - Pilgrim i
SUBJECT:
FACILITY TOUR FINDINGS, DECEMBER 8, 1987 facility tour on items on the attachment were noted during the The December 8, 1937. Please contact the Resident Inspector Office when your staff 4 is ready to discuss the evaluation of the items and the status of any actions taken. Please note the items and the facility response will be addressed in a routine inspection report. j Thank you for your time and attention to these matters. Sincerely, r e i Clay C. Warren Senior Resident Inspector ; l 1
Attachment:
As stated cc w/ attachment: R. Blough W. Kane W. Russell J. Wiggins i i l l
\
ATTACHMENT
-- Numerous motors appear to have f ailed grease seals caused by overgreasing without first removing grease drains. This condition causes a buildup of grease and dirt in the cooling airflow path and in extreme cases grease in the motor windings. (SBGT fans and SLC pumps) -- Nuts and bolts were noted laying inside an electrical cabinet in the RCIC room. -- Multiple cases of open junction boxes, terminal boxes and conduit pulled away from terminal boxes were noted.
Motor beaters for the "B" RHR pump appear to have overheated causing the insulation on the heaters to melt.
-- HPCI room cooler drip pan is full of paint scrappings which could lead to drain clogging. -- Stand 0y Liquid Control system relief valves have boric acid crystal buildup which could alter setpoints. -- Painting effort shocid be more closely controlled to prevent painting inaporopriate surfacts, i.e., linkages, valve packing glands, trip throttle valves, linit switches, etc. -- Numerous instances of scaffolding materials, i.e. , nails and wood chips, laying on floors. This caterial could migrate to drain systems and cause pu p or valve damage. Scaf folding was also noted attached to permanent equipment such as piping and conduit. -- Valve 1001-36A meter operator conduit had melted plastic cover.
l I l i 1 1 1 J
QUESTION 13. Has Pilgrim ever violated established radiation emission levels; i.e., have there been any releases from the plant which exceeded standards set by the NRC? ANSWER. The permissible levels of radiation in unrestricted areas and cf radioactivity , in effluents to unrestricted areas are established in NRC regulations embodied in 10 CFR Part 20, Standards for Protection Against Radiation. These regulations specify limits on levels of radiation and limits on concentrations of radio-nuclides in the facility's effluent releases to the air and water (above natural background) under which the reactor must operate. Further, the regulations require that there be no unmonitored release paths from the plant. The regulations are structured to provide reasonable assurance that no member of the general public in unrestricted areas will receive a radiation dose, as a result of facility operation, of more than 0.5 rem in 1 calendar year. These radiation-dose limits are established to protect the health and safety of the public. In addition to the Radiation Protection Standards of 10 CFR 20,10 CFR 50.36a establishes license requirements in the form of license Technical Specifica-tions on effluents from nuclear power reactors. The purpose of the Technical Specifications on effluents is to keep releases of radioactive materials to unrestricted areas during normal operations, including expected operational i I 4
l (Continued) 2 00ESTION 13. occurrences, as low as is reasonably achievable (ALARA). Appendix ! of l 10 CFR Part 50 provides numerical guidance on dose-design objectives for light l water reactors to meet this ALARA requirement. The dose-design objectives are j l low, about 1% of the Radiation Protection Standards of 10 CFR Part 20. Thus, it is possible for a licensee to exceed the dose-design ob.iectives, but still be l within the Radiation Protection Standards. ! I The NRC staff has reviewed the agency records on radioactivity releases from the Pilgtim nuclear power plant. Although there were situations when the radioactivity releases exceeded Pilgrim's Technical Specifications, these releases did not exceed the Radiation Protection Standards of 10 CFR Part 20. We have also reviewed the agency records on the amounts of radioactivity measured in the environment around the Pilgrim nuclear power plant. The licensee has reported elevated levels above normal background of some radionuclides in some environmental samples over the time period 1978 through 1981. However, it should be noted that Pilgrim's previous guidelines for reporting elevated levels of radioactivity in environmental samples were conservative. Under Pilgrim's current Technical Specifications, many (if not all) of the previously reported elevated levels would no longer be considered reportable. The previously reported elevated levels of radioactivity in environmental samples would lead to doses less than specified in the Radiation Protection Standards and thus would be below NRC regulatory limits. 1 1 I l i
NRC FORM 8 (4-79) NRCM O240 INCOMING AND SIGNATURE TAB
, Use this side of the sheet to precede the incoming material when assembling correspondcnce.
(USE REVERSE SIDE FOR SIGNATURE TAB)
NRC F@RM 8 (4-79) NRCM 0240 INCOMING AND SIGNATURE TAB Use this side of the sheet to precede the signature page when assembling correspondence. (USE REVER5E SIDE FOR INCOMING TAB)
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*., UNITED STATES
,8 g NUCLEAR REGULATORY COMMISSION 5 j W ASHINGTON, D. C. 20$55
*s ...../ .
EDO PRINCIPAL CORRESPONDENCE CONTROL FROM: DUE: 03/25/88 EDO CONTROL: 003569 DOC DT: 03/08/88 CEN. EDWARD M. KENNEDY FINAL REPLY: TO: MURLEY FOR SIGNATURE OF ** PRIORITY ** SECY NO: DESC: ROUTING: O'S FROM 1/7/88 HEARING ON PROPOSED RESTART OF STELLO PILGRIM TAYLOR REHM DATE: 03/14/88 RUSSELL ASSIGNED TO: NRR CONTACT: MURLEY MURRAY THOMPSON BECKJORD CPECIAL INSTRUCTIONS OR REMARKS: OCA SECY XDSEND VIA 5520 TO EDO IN Q&A FORMAT. FORMAT ATTACHED. b
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OFFICE OF THE SECRETARY CORRESPONDENCE CONTROL TICKET PAPER NUMBER: CRC-88-0220 LOGGING DATE: Mar 16 88 ACTION OFFICE: EDO AUTHOR: E.M. Kennedy AFFILIATION: U.S. SENATE LETTER DATE: Mar 8 88 FILE CODE: ID&R-5 Pilgrim
SUBJECT:
Questions re the proposed restart of Pilgrim ACTION: Appropriate DISTRIBUTION: RF, OCA, Docket SPECIAL MANDLING: None NOTES: DATE DUE: SIGNATURE: . DATE SIGNED: AFFILIATION:
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Rec'd off. @Q ' Date 3 5 I f -f T_ Time - Tk l}}