IR 05000219/1979010
| ML19260A818 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 08/22/1979 |
| From: | Briggs L, Neely D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19260A803 | List: |
| References | |
| 50-219-79-10, NUDOCS 7912030256 | |
| Download: ML19260A818 (20) | |
Text
{{#Wiki_filter:. . U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I , Report No. 50-219/79-10 Docket No.
50-219 License No.
OPR-16 Priority -- Category C Licensee: Jersey Central Power and Light Company Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960 Facility Name: Oyster Creek Unit 1 Inspection at: Forked River, New Jersey Inspection conduc.ced: May 3-4, 1979 lwy84g fj7/7ey Inspectors: L.BNg, Re -tf' Inspector date ' sign &d / ( A s@bg - D./heely,1dacto I sp8ctor date signed h$
G.'KalmaCReactorInspector date signed Approved by: 0 6 k 0+ t*,h fI s 7 /19 - E. C. McCabe, Jr., Chief, Reactor Projects date signed Section No. 2, RO&NS Branch Inspection Summary: Inspection on May 3-4, 1979 (Report No. 50-219/79-10) Special inspection of the low-low-low level condition occurring after the reactor trip on 5/2/79, involving 117 hours ensite by 4 region-based inspectors, one regional section chief, the regional director, the regional public affairs officer, and 4 NRR representatives.
Results: One item of noncompliance was identified (Violation: Deficient procedural controls result:d in isolation of all recirculatic;. loops with the consequent dropping of core water level below 7 ft. 2 in. without required core spray initiation, dropping of core water level below the 4 ft. 8 in.
Safety Limit for the Shutdown Mode, and starting of pumps in 2 idle recir-culatlan loops with temperature difference above the 50 F thermal limit esta-blished Detween the loops and the reactor coolant. Detail 4.b) 1461 155 Region I Form 12 (Rev. April 77) 'D12030-j f f(o '
. . DETAILS 1.
Persons Contacted
- J. T. Carroll, Jr., Station Superintendent K. O. E. Fickeissen, Technical Engineer
- I. R. Finfrock, Jr., Vice President - Generation E. Growney, Operations Engineer
- D. A. Ross, Manager - Nuclear Generating Stations J. L. Sullivan, Chief Engineer Other Accompanying Personnel S. Nowicki, Operating Reactors Branch 2, NRR R. Woods, Reactor Safety Branch, NRR R. Frahm, Reactor Systems Branch, NRR C. Berlinger, Reactor Safety Branch, NRR Other licensee representatives were contacted during the inspection.
These included reactor operations personnel, shift supervision, union representative, health physics personnel, security personnel, and members of the plant technical staff.
- denotes those present at exit interview.
2.
Logs and Records The following logs and records were reviewed without comment for the intervals as referenced except as noted elsewhere in this report.
a.
Station Log Book, May 2, 1979 b.
Drily Log, May 2,1979
- .
Events Recorder Traces, May 2, 1979 d.
Computer Printouts, May 2, 1979 e.
Recorder Charts, stack, offgas, pressure, level, and recirculation loop flows May 2, 1979 f.
Various operating procedures as referenced within this report.
3.
Circumstances Surrounding Reactor Trip of May 2, 1979 a.
Initiating Event Following surveillance testing of pressure switches which are associated with Isolation Condenser Automatic Initiation and Recirculation Pump Trips, an Instrument Technician proceeded to verify that the sensing line 1461 156
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. excess flow check valve was open.
During this verification, manipu-lation of manual valves on the sensing line caused a pressure spike affecting pressure switches attached to the line.
The pressure spike was sufficient to cause a high pressure reactor scram nd recircula-tion pump trip.
The acceptability of the licensee's procedures for verification of excess flow check valve position following instrument surveillance is an unresolved item (79-10-01) pending further NRC review of procedural changes in this area.
During the sequence of events that followed, (discussed in subsequent paragraphs), a Triple Low Water Level Alarm was received.
Upon the inspectors arrival onsite, the reactor was determined to be in the cold shutdown condi-tion with the MSIVs closed.
The inspector reviewed stack gas and offgas recorder records in the control room and verified that no abnormal offgas or stack gas conditions existed or had existed during the transient.
The control room log indicated that the mode switch had been placed in shutdown at 10:28 p.m. on May 2, 1979.
b.
Probable Sequence of Events The following sequence of events was reconstructed based upon recorder chart reviews, the events recorder, process printouts and discussion with operations personnel.
On May 2, 1979, at 1:51 p.m. a reactor scram occurred during required surveillance testing on the isolation condenser attributed to a momentary operation of two of the four reactor high pressure scram sensors which share a common sensing line with the isolation condenser high pressure switches being tested.
The scram occurred when the technician performing the test was verifying that the sensing line excess flow check valve was open.
TIME (Seconds) EVENT
REACTOR SCRAM /RECIRC. PUMP TRIP
TURBINE TRIP / LOSS OF FEEDWATER/ CONDENSATE PUMPS 13.6 LOW WATER LEVEL SCRAM POINT 16.8
- 2 RPS MOTOR GEN. SET TRIP
- 2 DIESEL GENERATOR BREAKER CLOSED
MSIVs CLOSED
B 150. CONDENSER IN SERVICE
LOW WATER LEVEL ALARM CLEARED 1461 157
. . .
172 LOW LOW LOW WATER LEVEL TRIP POINT 186 RECIRC. LOOP DISCHARGE VALVES FULLY CLOSED 1914 C RECIRC. PUMP STARTED 2208 FEEDWATER PUMP STARTED 2340 A RECIRC. PUMP STARTED 2700 RPS #2 RESTARTED / SCRAM RESET 3600 SB (BANK 6) RETURNED TO SERVICE The inspector's review determined that the events recorder capability was lost about 10 seconds after the low-low-low water level trip point was reached.
Loss of slow speed and/or fast speed capability was attributed to checks made to determine the initiating event.
The licensee committed to maintain the Events Recorder in an operable status in the future in order to provide retrievable information in analyzing transient conditions. This item is unresolved pending NRC review of the licensee's action to assure operability. (79-10-02) c.
Major Causative and/or Contributing Factors Inspector's review of the loss of feedwater transient which occurred on May 2,1979, identifies the following factors which caused or contributed to specific events culminating in a valid low-low-low level indication as measured inside the shroud.
(1) Standing Order No. 23 " Isolation Condenser iperation" dated November 15, 1977, required tripping of both "A' and "E" Recircu-lation Pumps, opening of the associated two U) inch bypass valves and simultaneous closure of the ay,clated discharge valves.
This action was previously required to preclude automatic isolation of the isolation condensers due to high differential ;1ressure caused by both the flow of cold water and recirculation loop flow contribution to differential pressure.
In that an automatic trip of all operating recirculation pumps now occurs '.s a result of the ATWS modification coincident with automatic initiation of the isolation condensers, simultaneous closure of the "A" and "E" valves is not now required to preclude improper isolation of the isolation condensers.
Deletion or changes to this Standing Order including conditions when recirculation pumps are not tripped are considered unresolved pending NRC review of such changes and licensee training.
(79-10-03) 1461 158
. . .
(2) Closure of the remaining open recirculation pump discharge valves in anticipation of pump restart (reference Detail 4.a regarding inadequate guidance and instruction to operations personnel concerning loop isolation).
(3) Control room indicatoas of reactor vessel level define the level in the annulus region and provided a misleading indication of normal reactor water level to the operators.
The low-low-low level instrumentation with taps located at the core spray spargers provides a level measurement inside the shroud and alarms in the control room but does not provide the control room operators with a level measurement for the core region.
(Reference Detail 5) (4) The absence of a low-low alarm signal and the clearing of the low level scram point alarm also provided misleading indications of proper water level to the operators. (Reference Detail 5) (5) Feedwater capability was reduced because one of the two startup transformers was out of service for inspection of its 4160V cabling. When the "A" Enawater Pump tripped and could not be restarted because of an auxiliary oil pump problem, no other feed pumps were available as a source of make up water to the reactor vessel-The time frame (36 minutes) to restore the "A" Feedwater Pump to service is unresolved pending further NRC review.
(79-10-04) d.
Licensee Commitment Concerning Restart The licensee stated that the transient would be treated as if a safety limit had been exceeded and that the plant would not be restarted until the NRC had reviewed the licensee's evaluation of the transient and granted approval for startup.
4.
Primary Coolant System a.
Recirculation Pump Discharge Valve Closure The following procedures and standing orders were reviewed to ascertain requirements placed upon operations personnel regarding recirculation pump valve manipulations: Standing Order No. 23, Revision 0, dated November 15, 1977.
-- -- Procedure No. 301 " Nuclear Steam Supply System," Revision 7, dated September 22, 1978.
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No. 532, Automatic and Manual Scram, Revision 1, March 28,1977.
-- No. 511.1, Feedwater Pump Failure, Revision 1, July 5,1978.
-- -- No. 514, Reactor Isolation Scram, Revision 2, November 12, 1976.
Prior to the transient of May 2,1979, the "D" recirculation pump had been placed out of service with its discharge valve closed and its suction and discharge bypass valves open. The remaining four loops were in operation with pumps running, suction and discharge valves open, and discharge bypass valves open. As a result of the initiating event, all operating recirculation pumps tripped as designed.
seconds into the transient, the "B" Isolation condenser was placed in service, and the "A" and "E" recirculation pump discharge valves were closed. Standing Order No. 23 states in part that: "Whenever the conditions exist following a Reactor Scram that require initiation of the isolation condensers, perform the following additional steps: (1) Trip off both the ' A" and 'E' Recirc Pumps (2) Open the associated bypass valves (3) Simultaneously close the associated discharge valves.
The Standing Order indicates these actions are taken to preclude possible isolation of the isolation condensers due to an apparent line break (high flow condition).
No further valve manipulations are required by the procedures. The sequence of events indicates that the "B" and "C" recirculation loop discharge valves w re closed.
This was apparently done to satisfy recirculation pump start interlocks in anticipation of restarting the pumps. Procedure No. 301 " Nuclear Steam Supply System," Revision 7, September 22, 1978, states in part: Never isolate all recirc pumps with fuel in the reactor and the reactor head in place." The interviews indicate that " isolated" was interpreted by the operators to mean that both the discharge and discharge bypass valves are shut.
The adequacy of this procedure to accomplish its intent and provide sufficient operator guidance is considered an unresolved item (79-10-05) pending changes as well at operator training in this area.
b.
Review of the Emergency Procedures, noted above, as compared to Procedure No. 301 show that the Emergency Procedures did not incorporate a suitable restriction on Recirculation Loop Isolation to insure that operations personnel were fully aware of level differences which could develop if communications between the core and annulus region was restricted by the closure of all loop suction and/or discharge valves. The emergency procedures that were used to cope with this event are considered not appropriate for the circumstances as defined in 10 CFR 1461 160
. .
Part 50, Appendix B, Criterion 5, and is considered to be a major contributing factor in the May 2,1979 event. This item is considered a noncomplianu. (79-10-13) c.
Recirculation Pumo Restarts _ Later in the transient the "C" and "A" recirculation pumps were restarted (at 1914 and 2340 seconds, respectively). The "C" pump was stopped about 2 minutes later because of an excessive drop in indicated water level. The feedwater system was returned to service and reestablished water level and the "A" recirculation pump was then started and remained in service.
On each restart, the recirculation loop coolant to reactor coolant differential temperatures exceeded the Technical Specification Limiting Condition for Operation of 50 F for restart of an idle recirculation loop.
The actual differential temperatures were 72 F and 92 F for "C" and "A" loops respectively. This item is included in the Notice of Violation as a consequence of the event.
5.
Level Instrumentation a.
Description Level instrumentation at the Oyster Creek station consists of four ranges of reactor vessel water level measurement.
Each range is monitored and indicated as outlined below. Although each range has its own reference point, all vessel components are referenced to height in inches above the inside invert of the vessel bottom head.
(The top of the active fuel is at 353".)
A single wide range instrument (GEMAC) monitors vessel annulus level from a reference point of 443" to above the vessel top head. This instrument is used for indication only and is scaled to read 0-400", with 0" at vessel level 443".
Indication is available only in the control room.
Two narrow range instruments (GEMAC) monitor level from a reference point of 443" vessel annulus level to a full scale reading of 539" vessel level. Each of these instruments is scaled to read from 0'-8' (0'=443" vessel level) and uses separate instrument taps on the reactor vessel. These instruments are used in the Feedwater Control System, each is indicated and one is recorded (selectable) in the control room. No local in plant indication is available.
1461 161
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Eight narrow range instruments (YARWAY) monitor vessel annulus level from 439" to 539".
Each of these instruments is scaled to read 0"- 100" (0"=439" vessel level) and one half use separate vessel instru-ment taps. These instruments are used for protective features and for low and low-low level alarms.
All eight are indicated locally in-plant, and two are indicated in the control room.
Four wide range instruments (BARTON) monitor vessel core region level from 409" - 541".
Each of these instruments is scaled to read 0" - 150" (0=409" vessel level); two use one core spray system vessel penetration as a variable leg and one narrow range (GEMAC) upper instrument tap as a reference leg. These instruments are used in ADS logic and for the low-low-low level alarms in the control room; all are indicated locally in plant, with no control room level measurement indication.
b.
Perforraance During Transient on May 2,1979 Since the level taps for the wide range (GEMAC) and narrow range (GEMAC and YARWAY) instruments penetrate only the reactor vessel wall, these instruments monitor only the level in the annulus region outside the core shroud.
The wide range (BARTON) instruments, however, monitor level inside the core shroud via the core spray spargers.
During the transient, the reactor core region water level (inside the shroud) was being reduced because the isolation condensers draw steam from the core region. Water level outside the shroud was being main-tained or increased by the condensate return from the isolation condensers.
The misleading information presented to the operators was an indication of normal level for the existing conditions.
Conflicting information was presented when a low-low-low alarm was received with no low-low level alarm present.
The conflicting level alarm indication is considered an unresolved item (79-10-07) pending operator training on level instrumentation and further NRC review. The unavailability of the wide range (BARTON) indication in the control room for operator assessment of level is considered an unresolved item (79-10-08) pending licensee management disposition and further NRC review.
c.
Level Indication Methods As described in Detail 5.a, there are three specific vessel levels which are indicated on instruments as 0 indication. In addition, two scaling methods for indication are used; one in inches of level, the other in feet and inches of level. Both of these items complicate correlation of different level indicators by operators during a transient.
This area is considered unresolved (79-10-09) pending licensee disposition and NRC review.
1461 162
.
d.
Reactor Low-Low Low Water Test and Calibration The inspector reviewed the following test and calibration data to determine relative status of the low-low-low level instrumentation prior to the reactor trip on May 2,1979.
(1) Surveillance procedure "Rx Triple Low Water Level Test and Calibration," Revision 1, April 3, 1978, and results were reviewed for the following dates: 2/14/79 test and calibration 3/15/79 test only 4/12/79 test only This procedure documents the calibration of sensors RE18A - D and/or documents the testing of the triple low water level trip including setpoint, vessel low-low-low water level annunciator, event recorder response, and excess flow check valve position.
(2) The setpoint for the low-low-low water level trip is +10" on sensors RE18A - D based on the vessel water level Safety Limit of 4'8" above the top of the active fuel with the reactor shutdown, instrument error, and density corrections applied.
These sensors utilize the core spray spargers which are located 4'8" above the top of the active fuel as a variable leg tap. This level is also designated as instrument zero.
The instrument error is +3".
The density correction consists of two terms: 1) the difference in the setpoint resulting from the density of water at calibration temperatures and at maximum reference leg temperature; 2) the weight of steam existing between the reference and variable leg. As calculated by the licensee these corrections are a maximum of 1.75" and 4.8" respectively, and are applied as a +7" correction. The acceptability of the correction factors with respect to the 4 foot 8 inch limit is considered unresolved pending further NRC review. (79-10-10) (3) The inspector's review of these surveillance procedures revealed no unacceptable conditions in sensors RE18A - O performance.
6.
Health Physics Details of the Health Physics section of this report will be contained in IE Inspection Report No. 50-219/W13.
1461 163
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7.
Environmental Radiation Monitoring a.
Releases The inspector examined the plant releases during and after the reported plant transient on May 2, 1979. The plant stack gas samples which were collected before and during the transient showed essentially the same concentrations of radioactive and other nuclides (Table 1).
During the transient and the Cooldown phase, an approximate 107,500 gallons of water was released as steam through the Isolation Condenser vent. The make up water for the Isolation Condenser was provided from the Condensate Storage Tank. The inspector collected water samples from the Isolation Condenser and the Condensate Storage Tank. The analytical results of these samples are in Table 2.
The 4 npector examined the licensee's calculations for the total activity released with the vented steam and verified by independent sampling, analyses and calculations that < 10 mci, < 3 mci and < 0.5 mci of Co-60, Mn-54 and I-131, respectively, were released, b.
Onsite Sampling and Analyses The inspector examined the site cad noted chromated water puddles outside the reactor and radwaste buildings. The licensee stated that the observed water puddles resulted from the condensation of the vented steam from the Isolation Condenser. This water was sampled and analyzed by the NRC. The analytical results showed low levels of Co-60 and Mn-54 activity (Table 2). The inspector determined that the contaminated chromated water was confined to the immediate onsite area near the Isolation Condenser vent, c.
Environmental Sampling and Analysis Independent sampling and analysis of environmental media were performed by the NRC. Based on the meteorological conditions during the transient (Table 3) and the X/Q values for offsite locations at distances between 500 and 3000 meters from the plant, the inspector determined the appropriate offsite sampling locations.
Several water, grass, soil and vegetation samples were collected f om these locations. These samples were analyzed by the NRC Mobile Laboratory and showed no evidence of radioactive contamination (Table 4).
Direct radiation measurements were made by the licensee using thermo_- luminescent dosimeters (TLDs) and film badges.
A sample of the offsite TLDs was read and' the preliminary results are shown in Table 5 and Figure 1.
The data samples were taken from locations surrounding the plant and indicate that readings are not attributable to abnormal plant operations since in all readings regardless of wind direction - are comparable. The licensee's environmental air samples analysis is still in progress.
1461 164
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8.
Preliminary NRC Determinations Concerning Loss of Feedwater Transient of May 2, 1979 Four NRR representatives arrived onsite Ma 3, 1979 to review the transient.
Areas examined are discussed below: a.
Minimum Water Level Experienced in the Reactor Vessel The NRC team onsite provided NRC Headquarters with data for performing a preliminary calculation to determine the minimum water level of the reactor core during the transient.
The General Electric Company per-formed a similar analysis.
The results of GE's analysis were discussed with the NRR team.
b.
Possible Uncovering of Reactor Core The NRR team reviewed pressure and temperature records, sequence of events records, process printouts, offgas and stack gas records, and held discussions with members of the plant staff to determine if the reactor core was uncovered during the transient.
c.
Indications of Fuel Damaae The NRC onsite team reviewed stack gas racords and reactor coolant samp;a analyses to determine if indications of fuel damage from the transient were present.
Effluent records were normal.
d.
Cause of the Transient Review of data and interviews with operators confirmed the reported sequence of events, and determined that enough data was available to determine why this transient occurred.
e.
Conclusions Based on analysis of data and interviews, the NRR staff's preliminary conclusions were that (1) the fuel had not been uncovered during the transient; (2) there were no apparent indications of fuel damage; and, (3) enough data were available to ascertain why the transient progressed the way it did.
The licensee's submittal to NRR had not been made at the conclusion of this inspection.
This item (79-10-12) is considered unresolved pending NRC review of the licensee's evaluation of the transient and inventory calculations and assumptions including a submittal of this information to the NRC.
1461 165
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9.
Unresolved Items Areas for which additional information is required to determine acceptability are considered unreselved.
Unresolved items identified are contained in Details 3, 4, 5, and 8.
10.
Exit Interview An exit interview was conducted at the conclusion of the inspection on May 4, 1979 with senior management representatives denoted in Detail 1.
The licensee acknowledged inspection findings.
1461 166
. Table 2 NRC Analytical Results of In Plant Samples Collected on 5/5/1979 at Oyster Creek Nuclear Generating Station Concentration uCi/ml Sample Description I-131 Cs-137 Mn-54 Co-60 Others 7'olation Condenser <6.34E-7 <8.31E-7 1.83E-5 5.65E-5 Later Shell Side "A" 1433 hrs.
Isolation Condenser <6.34E-7 <8.31E-7 <7.6E-7 7.99E-6 Water Shell Side "B" 1438 hrs.
Condensate Storage 1.15E-6 <8.31E-7 <3.64E-7 2.23E-6 Ce 141 1446 hrs.
3.92E-6 Water on Top of <6.34E-7 <8.31E-7 <7.64E-7 <8.85E-7 Empty drums outside Radwaste Office Trailer at a 300 ft. N.E. of Reactor Bld. 1545 hrs.
Water puddle under <6.34E-7 <8.31E-7 1.15E-5 3.23E-5 Isolation Condenser Vent 1545 hrs.
1461 167
. TABLE 3 Meteorological Conditions and X/Q values For Ground Level Release at Oyster Creek Nuclear Generating Station on 5/2/1979 Wind Wind Direction 500 1000 3000 Time Speed From a Temp.
meters meters meters 12 noon 8.0 W-2.6 9.34 E-6 1.07 E-6 1.89 E-7 1300 9.1 WNW-2.4 8.21 E-6 9.41 E-7 1.66 E-7 1400 7.4 WNW-2.4 1.01 E-5 1.16 E-6 2.05 E-7 1500 7.1 WNW-2.2 1.05 E-5 1.21 E-6 2.13 E-7 1600 10.3 SSE-2.2 7.25 E-6 8.31 E-7 1.47 E-7 1700 8.2 S-1.7 9.11 E-6 1.04 E-6 1.85 E-7 1800 5.1 S-1.4 1.46 E-5 1.68 E-6 2.97 E-7 1900 3.5 S-0.5 9.54 E-5 3.46 E-5 6.11 E-6 2000 4.1 WSW +0.7 9.76 E-5 3.85 E-5 7.75 E-6 2100 5.6 WSW +1.8 1.13 E-4 3.49 E-5 8.53 E-6 2200 6.3 WSW +1.6 1.00 E-4 3.11 E-5 7.68 E-6 2300 5.6 SW +1.3 1.13 E-4 3.49 E-5 8.53 E-6 2400 5.8 WSW +0.6 6.90 E-5 2.71 E-5 5.48 E-6 146i l68
. . . TABLE 4 NRC Analytical Results of Environmental Samples Collected on 5/5/1979 Oyster Creek Nuclear Generating Station Concentration uCi/ml Sample Description I-131 Cs-137 Mn-54 Co-60 Others Standing Water 0.7 <6.3.4E-7 <8.31E-7 <7.64E-7 <8.85E-7 mile' North of the plant Stream Water 0.7 <6.34E-7 <8.31E-7 <7.6E-7 <8.85E-7 mile-North of the plant
Soil Im ; 1/2" deep <l.69E-7 <2.5E-7 <2.37E-7 <3.04E.7 1 mile-North of the plant
Soil 1m ; 1/2" deen <l.69E-7 <2.5E-7 <2.37E-7 <3.04E.7 0.5 mile; East South East of the plant
Soil Im ; 1/2" deep <l.69E-7 <2.5E-7 <2.37E-7 3.04E-7 0.7 mile; North of the plant
Soil 1m ; 1/2" deep <1.69E-7 <2.5E-7 <2.37E-7 <3.04E-7 0.8 mile; South of the plant
Grass (lm ) <8.44E-7 <1.25E-4 <l.19E-4 <l.52E-4 0.7 mile; North of the plant
Grass (lm ) 1.0 mile; <8.44E-7 <l.25E-4 <l.19F-4 <l.52E-4 North of the Plant
Vegetation (lm ) 0.8 <8.44E-7 <l.25E-4 <l.19E-4 <l.52E-4 mile; South of the Plant
Grass (lm ) 1.0 mile; <8.44E-7 <1.25E-4 <l.19E-4 < f2f 4 East South East of the plant Vegetation 0.5 mile, <8.44E-7 <1.25E-4 <1. l N-4 < i. t:2 E-4 East South East of the plant )4h} }6h
. . . . TABLE 5 Environmental Direct Radiation Measurements (TLD) Oyster Creek Nuclear Generating Station Exposure Period m rads Location Post Transient Station 2 May 2 - May 6 2.5 1 0.26 Station 6 May 2 - May 6 2.0 1 0.53 Station 7 May 2 - May 6 2.53 1 0.33 Station 8 April 30 - May 6 2.32 1 0.23 Station 9 April 30 - May 6 2.57 1 0.41 Station 10 May 1 - May 6 2.58 1 0.25 Station 13 May 1 - May 6 2.05 1 0.38 1461 170 _
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Isotope Pro-Gross Bkg.
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2 l35
Gross Counts / min.
g*O [328[ 2,
f,6 (f - f A,2.ta f "' I l35 I
51
-' Bkg. Counts / min.
. /, 4 j 7,4 Cr
Cr . Mn
Mn
Net Counts / min.
7,. [ g,' f,
54' " /g g j.g[8 CoSO
/, }f F #/ j '/. 7/F~ # 3 Co
Counter Factor
Net Die / min.
[.ggf QK8/[[ Fe
Fe59
Co
Co
/, 0,3 g -# / e/,f7p'~ / 4.5E-7 h frCi /.ffl.E-d' /, O $' E-[
60 f, '/ E$ '/ 7 Ch /W 6~ #
91 .4 j. 4 7 g - e 2.
Volume 'cc Sr
Sr
r
10 . zr
Conc.
mci /cc 7./h E7'p 3. f Uf'
zr
95
, /j Analysis Data Gross y Notifying Nb
Nb
9
Mo
Counter /[C" Mo
- ic , Tc99m
' Tc
Count Date & Time [-y ()pfe 99" 134 f9/'2 8 Gross a Cs
Cs
Gross Counts l37 137
Total Count Time
- /O min.
~ i Cs
Cs l<10 140 1G /,0 ? g -/# ,4 /,/75 Gross Counts / min.
Ifl7*h Gross 6 i -# 1 Da
Ba ., La
g3
g g g -// y,j,j.f f -s3 Dkg. Counts / min.
/ E[abe O 140 140 ~ _ l41 141 Not Counts / min.
~{78 $ O Il31 i A Ce
Co
- Ce
Ce
cpm /cc /* MO E-d' ! 144 239 ~## i Np
/,4t/ E A. J. 4 f / ~# 3 239 P,4I8 '# /. A t'# ~# 3emarks [ ~ Np
N I N ' OYSTER CREEK DATA b t 4t . . , _ _ _ _ _ _ _ _. _ _ _... _ -. ... .. - -. _.. _ .- --- - -- s.. para yye w w p.,. , - s a 9 W AV##AN %,C, , fM .. - ue v.wg,. ..-e,, %g-W,. ,, y- %y , ~ - ... lk.V{g,..sW$,h'135bh,l,.V
- - '
-t' fg.
fis- -
- -'
o ...... EM: p,h % ~? r.G' ,_ % ~ h'~- hihg-E v--- em.
a . m,+ e. f'gj n* y
.5 2b > ~ ' zn _ _ I . ..- a n a 11 ~. n'-,V .-,.-y'.,. %-i .. % _ .y,4 4 y.,
- -
T I - .. >. t.
_ 'E - (r.
r g' ~c h {^.l N . . F f -
'/ ? %. ~. &z.NO e W . ?_M j'r. --kj%~ !It' - ,.. ". __' k
' . =,.} : f"
f ^ -a . 1*h ' nl-
. P Writ =f ~ 5']=l6',.:;:; . k' ' V. L[+ M Y.r_&. . .. ~. - s 'i
y ' . -
. 4-h , 30 m - T [M... M = ' t ' ' d4 ' "
= ' "s "ay ?- : %,e .,] w ,
' / . ..f m.l~=h $ { u"' _ f'h _ ,
. , ^ e.. m ' '
- a g . s fr w a.z y,s - y :q, ,s'-J ...y - .
.,m.
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u e.r-I l { .$?S ' i.
=..-:a .J i -*
,#' ' . e
. ., b . ' . i ph"D '$k-a' 2._"sid=, ' w;g s.a ce,
4- '. (%~9Q % ,f b.- - ' &_c h.s '.a,' % $.?1 %. [V..w 2 > %$Nb :b \\t {uMh.f.h,. -,Q ' f'$y 'g,. - ** g f[ :.. u r.a - . ,
- eo A.my q" N.:. ff.:,.ers r- , i.
, ,1" *v l \\ ? f _~~~-NR
- . q1-W / 8'*a' 8'"a t'w i o S re.E - %rJE-h * 9 s_ w , ... g _ nm a ?h 1. D TN.
&n ..' - L 4' [ ' '
, -R:M d . J.r = G F*;_=3 s / - j Gur-2$T3. ' ~2^E f.. W -=ff Q . M 'O.' .. hCt'. b -l W = r s 9 - T;, '
- '.6y # -. I.
~ " " "** _.
Radiation Survey Statiens . Oyster Creek Nuclear Electric Generating Statica ~ 1461
5 D**}D *]D' ] oJu SU.kinlo 'l o o Ju Figure 1 L ... _ . . ~ - - -ea, e _ }}