IR 05000213/1981020
| ML20039E341 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck, Oyster Creek File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 11/06/1981 |
| From: | Robert Carlson, Roth J, Rich Smith NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20039E323 | List: |
| References | |
| 50-213-81-20, 50-219-81-38, NUDOCS 8201070171 | |
| Download: ML20039E341 (97) | |
Text
{{#Wiki_filter:_ _-_ , - - - - - - - _ - - - -. _ _ - -. - - -_ - _ -. - - - - ( I i I I U.S. NUCLEAR REGULATORY COMMISSION f 0FFICE OF INSPECTION AND ENFORCEMENT (
REGION I
, l l 50-213/80-20
Report No.
S0-219/80-38 50-213
Docket No.
50-219 ' DPR-61 License No.
DPR-16 Priority Category C -- Licensee: Connecticut Yankee Atenic Power Company Jersey Central Power and Light Company P.O. Box 270 Madison Avenue at Punch Bowl Road l Hartford, Connecticut 06101 Morristown, New Jersey 07960 Facility Name: Haddam Neck Plant Oyster Creek Nuclear Generating Station ' A Investigation at: Haddam Connecticut; Forked River, New Jersey; and Atlanta, Georgia Investigation conducted: From October 6, 1980 to January 14, 1981 Investigators: y Meu //J , f)ay ond H. Sm1th,'In'vestigator~' a)6 gned' t-% l0
/ ' ome' Roth, Fuel Facility Inspector ddte' signed date signed / 7 date signad _ Approved by: M // 4 8 / m , Robert T. Carlson, Director /dat(esigned Enforcement and Investigation Staff Investigation Summary: Investigation from October 6,1980 to January 14, 1981 (Report No. 50-213/80-20 and 50-219/80-38) Investigation of circumstances surrounding the transportation and use of the Model No. NFS-4, Serial 2. NAC-lE cask shipped from Haddam, Connecticut on May 1, 1980 until it arrived at Camp Pendleton California on August 2,0,, 1980.
t 8201070171 011223- PDR ADOCK 05000213i G PDR) J
SUMMARY By memore:dum dated October 7,1980, IE:HQ' requested Region I to investigate the circumstances surrounding the transportation and u'se of the Model No.
NFS-4, Serial No. NAC-lE shipping container.
Inspections had been conducted by NRC Regions I, III, and V which related to radiation and contamination associated with the transport and use of the container.
The container was loaded.on April 26, 1980, at the Haddam Neck Plant, Haddam, Connecticut, which is operated by the Connecticut Yankee Atomic Power Company (CYAPCO). The shipment contained a fuel bundle having known fuel cladding failures. Because of problems with contamination continuing to be emitted from the container walls following decontamination, and radiation levels which nece'ssitated placing external shielding on the container transport vehicle, departure of the shipment from the site was delayed until May 1, 1980. An NRC Inspector observed portions oflthe cask decontamination activity and was present when the cask departed the site.
On May 2, 1980, the shipment arrived at the Battelle, Columbus Laboratories (Battelle) facility located at West Jefferson, Ohio. A release of radioactivity occurred when the container was opened which caused abnormal radiation and contamination levels in the facility.
The container was decontaminated after-it was unloaded and departed the facility on July 22, 1980. The radioactive release was reported to the NRC Region III office and an inspection was performed September 22-26, 1980.
Inspection Report 70-008/80-02 was issued.
On July 23, 1980 the container arrived at.the Oyster Creek Nuclear Generating Station located at Forked River, New Jersey and operated by the Jersey Central Power and Light Company (JCP&L). The container was scheduled to be used for transporting a fuel bundle to Battelle for examination and study. JCP&L surveyed the container and reported to the NRC that contamination levels exceeded the DOT and NRC reporting levels. The radiation survey of the cask trailer revealed that in one area under the trailer the radiation level exceeded the DOT limit. This was confirmed by an NRC Inspector on site. The-container l was unloaded and subsequently the container and transport trailer were decontam-inated externally by Battelle representatives.
Representatives of the Nuclear ' , Assurance Corporation (NAC), the owners of the cask, were also at the site.
NAC made arrangements to have the container transported to San Onofre and the cask departed Oyster Creek on August 15, 1980.
On August 16, 1980, the shipment stopped at Battelle in West Jefferson, Ohio and the cask lifting yoke was off loaded for storage.
On August 20, 1980, the container. arrived at San Onofre Unit 1 located at Camp Pendleton, California and operated by the Southern California Edison Company.
. Receiving surveys determined that the truck cab sleeper radiation level was 4.4 mrem /hr which exceeded the DOT limit.
The cask was unloaded for decontamination. The cask handling and use were examined during an inspection-performed September 22-26, 1980, by Region V.
Inspection Report No. 50-206/80-26 was issued.
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The transportation of the Model No. NFS-4, Serial No. NAC-1E cask described supra involved the transport of an irradiated fuel bundle only between the Haddam Neck Plant and the Battelle facility at West Jefferson, Ohio. The radiation levels at the external surfaces of the cask ouring the other transfers of the cask indicated that radioactive material remained in the cask after it was decontaminated at the Battelle facility and appeared to change positions within the cask during transport.
The film badge dosimetry results of the drivers involved with the cask transport indicated that the drivers did not exceed the allowable quarterly exposure for radiation workers.
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PURPOSE OF INVESTIGATION The purpose of this investigation was to determine the circumstances surrounding the transportation and use of the Model No. NFS-4, Serial No. NAC-lE cask shipped from the Haddam Neck Plant, Haddam, Connecticut on May 1, 1930, until it arrived at San Onofre Unit 1, Camp Pendleton, California on August 20, 1980.
BACKGROUND The Connecticut Yankee Atomic Power Company (CYAPCO) forwarded a letter to the Director, Office of Nuclear Materials Safety and Safeguards (NMSS) dated February 22, 1980 (See Exhibit 1), requesting a route approval for shipments of irradiated fuel assemblies.
The shipments were to be delivered to Batti.11e Columbus Laboratories (BCL) for examination to determine the cause of fuel failures.
On March 27, 1980, a conference telephone call was hela between representatives of CYAPC0 and NRC, Office of Nuclear Reactor Regulation regarding the movement of a spent fuel cask into the spent fuel pool.
(see Exhibit 2) By letter dated March 25, 1980, the Nuclear Assurance Corporation (NAC) provided CYAPCO with information relative to some of the items discussed during the March 27, 1980 telephone conference.
(See Exhibit 3) By letter dated March 31, 1980, CYAPC0 received the route approval for shipments which had been requested previously.
(See Exhibit 4) A letter dated April 15, 1980, to the NRC, Region I, office from CYAPC0 informed the NRC of the intent to make three shipments of irradiated fuel assemblies.
(See Exhibit 5) CYAPC0 forwarded a letter dated April 18, 1980, to the NRC-NRR containing information relative to the items discussed during the March 27, 1980 telephone conference. This informat'on was provided to support a proposed change to technical specifications to permit spent fuel cask movement over the spent fuel pool in order to complete the planned ship?.ents.
(See Exhibit 6) The NRC issued license amendoent No. 35 to License No. DPR-61 dated April 24, '980, which permitted CYAPC0 'o move the casks involved in the planned loading and shipping of irradiated fue, assemblies.
During April 28, 29 and 30, 1980 there were telephone discussions between representatives of CYAPCO, NRC Inspector at site, NRC Region I, NRC-IE:HQ, and NRC-NK aqarding the problems encountered after the cask, Serial No. NAC-lE, was loaden with irradiated fuel for shipment. The problems encountered were with " lea:hing,d or " weeping" of contamination frem the cask. This problem was requiring the licensee to decontaminate the cask several times and was a concern t9at the cask contamination levels could possibly exceed the Department of Transpc.-tatf or (DOT) limits during shipment.
There was also a problem of radiation hvels exceeding DOT limits.
The cask contamination was reduced - _ _ _ .
below DOT limits after several decontaminations and the radiation levels were reduced below DOT limits by the addition of external lead shielding around portions of the shipping cask.
The shipment (Licensee No. 0-80-10) departed from the Haddam Neck site at about 1:30 p.m. on May 1, 1980, enroute to Battelle Columbus Laboratories, West Jefferson, Ohio.
(Ref: Inspection Report No. 50-213/80-07) On May 16, 1980, NRC-Region I was notified by NRC-NMS5 that CYAPCO had reported that the decay heat content limits for the shipment of May 1, 1980, had been exceeded. Telephor.e discussions followed between CYAPCO, NRC-Region I, NMSS and NRC-IE:HQ. The licensee informed Region I that the calculations indicated a decay heat content of 2.09 Kw prior to the shipment and that subsequent calculations had determined a decay heat content of 3.5 Kw.
This information was confirmed by letter dated May 21, 1980 (See Exhibit 7).
The licensee also informed Region I that the shipment had arrived at Battelle on May 2, 1980.
Also on May 16, 1980, Region I contacted Battelle Columbus Laboratories by telephone and was informed that the contamination and radiation levels of the CYAPC0 shipment received on May 2, 1980 were within DOT limits.
The Battelle representative also indicated that no problems were encountered in handling the cask other than what was expected.
Since the shipment contained failed fuel rods the contamination level inside the cask was expected to be high.
The representative also stated that the pool water became contaminated when the cask was opened.
The Battelle Columbus Laboratories submitted a report, pursuant to 10 CFR 21, Reporting of Defects and Noncompliance, on June 27, 1980, to the NRC-Region III Office. This report described the release of radioactive material when the shipping cask containing the failed fuel assemoly was opened in the fuel pool (Ref: Inspection Report No. 70-008/80-02).
On July 23, 1980, the NAC-lE shipping container arrived at the Jersey Central Power and Light Company, (JCP&L) Oyster Creek Nuclear Generating Station located at Forked River, New Jersey. The container had been shipped empty from Battelle Columbus Laborate-ies, West Jefferson, Ohio. JCP&L refused to accept the shipment due to the radiation and contamination levels on the container. A radiation specialist from NRC-Region I was present at the Oyster Creek facility. The container was released from the Oyster Creek facility on August 15, 1960 (See Exhibit 8).
On July 24, 1980, the NRC-Region I issued Preliminary Notification No. 80-109, " Contaminated Empty Spent Fuel Shipping Cask" (See Exhibit 9).
On August 20, 1980, the empty container shipped from Oyster Creek arrived at San Onofre Unit i located at Camp Pendleton, California.
Radiation surveys conducted by the licensee (Southern California Edison Company) determined that the dose rate measured in the tractor sleeper was 4.4 mrem /hr which exceeded the DOT limit of 2.0 mrem /hr. The contamination levels of the shipment were within 00T limits (Ref: Inspection Report No. 50-206/80-26).
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DETAILS Spent Fuel Shipping Cask-Model No. NFS-4 A Certificate of Compliance No. 6698 was issued by the NRC for the above cask based on an application submitted by Nuclear Fuel Services, Inc., dated October 6, 1972. The certificate has been revised several times and the Safety Analysis Report has also been updated.
The revised certificates also reference submittals to the NRC from the Nuclear Assurance Corporation (NAC).
Revision 9 of Certificate No. 6698, dated December 13, 1979 was in effect when CYAPC0 made the shipment on May 1, 1980, with the Model No. NFS-4, Serial No. NAC-lE cask.
(See Exhibit 10) Shipments Made of Cask Model No. NFS-4, Serial No. NAC-1E Departed CYAPCO - May 1, 1980 Arrived Battelle - May 2, 1980 Departed Battelle - July 22, 1980 Arrived Oyster Creek - July 23, 1980
- Departed Oyster Creek
- August 15, 1980 Arrived San Onofre - August 20, 1980
- with a stop off at Battelle.
All cask movements were transported by Tri d tate Motor Transit Company.
Activities Involving Cask Model No. NFS-4, Serial No. NAC-1E, at CYAPCO As stated previously in the " Background" portion of this report, the NRC was aware of CYAPC0 planning to use the NAC-lE container and the contamination / radiatiot, problems which they encountered after it was loaded.
On October 10,16, and 17,1980, R. Smith, NRC Investigator was at the Haddam Neck Plant to review the circumstances surrounding the use of the container.
The following individuals were contacted:
- R. Begenski, Reactor Engineer H. Clow, Health Physics Supervisor R. Graves, Station Superintendent
- M. Hills, Supervisor, Reactor Performance Section W. Nevelos, Radioactive Waste Foreman A. Niriccio, Nuclear Information Supervisor
- M. Pitek, Staff Engineer, Reactor Engineering Branch D. Vement, Nuclear Records Supervisor
- By subsequent telephone discussions.
- Northeast Utilities Service Company (NUSCO)
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The investigator examined the following documents related to the use of the NAC-lE container which was shipment Nc. 0-80-10 (licensee number): Certificate of Compliance Safety Analysis Report Radiation Work Permits Radiation and Contamination Surveys Shipment Records Procedures for Unloading and Loading the NAC-1 Series Containers Based on an examination of documents and discussions with licensee representatives, the following information was developed: Certificate of Compliance No. 6698, Revision 9 was amended by an NRC order dated December 12, 1979. A list of registered users was attached to the order and CYAPC0 was not named as a registered user. CYAPC0 submitted a letter to the Director, Office of Nuclear Materials Safety and Safeguards dated February 22, 1980 (See Exhibit 1) which contained the information required by 10 CFR 71.12(b).
This letter and documentation examined at the site showed CYAPC0 to be a general licensee user of the container.
CYAPC0 had contracted with NAC to provide the NAC-1 shipping container and services. NAC had provided surveillance testing on the container at the site.
NAC had also provided oversight for the container loading and had two representatives at the site. During the surveillance testing, one of the drain ball valves would not provide a seal.
The ball valve was removed and replaced with a pipe plug.
(See Condition 9 of Exhibit 10, Certificate of Compliance).
After completion of the surveillance testing, the container was loaded on April 26, 1980.
The NAC-1E container was loaded without adding or removing any spacers in the container. The temperature was obtained on the final gallon of water drained from the loaded container to compare with the temperature of the water in the spent fuel pit. The check sheet record had no entry recorded for the temperature of the final gallon; a temperature of 85'F was recorded for the spent fuel pit; and, 76'F was recorded as the temperature rise.
R. Begenski, who performed the temperature measurements, stated that the 76'F figure was improperly recorded and was the temperature of the final gallon drained from the container. He also stated that in this shipment and two later shipments of spent fuel, the temperature of the final gallon drained from each loaded container was lower than the temperature of the water in the spent fuel pit. He also recalled that a NAC representa-tive was present during draining of the container and obtaining the temperature of the drain water.
The pipe plug that had been installed during the surveillance tests conducted by NAC was not removed after the cask was loaded or prior to departure from the Haddam Neck Plant.
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The difficulty in decontaminating the container was indicated by the Radiation Work Permits and the radiation survey data. CYAPC0 had contacted representatives of NAC and Battelle regarding the " weeping" problem of the container and was informed that the problem had previously been encountered during other shipments. The final contamination and radiation surveys prior to departure indicated that the levels were within D0T limits. Two representatives from CYAPC0 accompanied the shipment until
its arrival at Battelle.
(See Exhibit 11, Bill of Lading) No problems ' were encountered with the shipment and the surveys upon arrival were
within DOT limits.
Battelle contacted CYAPC0/NUSCO (Northeast Utilities Service Company) regarding the contamination problems-experienced when the container was opened in their fuel pool.
The calculations for the decay heat content reported to the NRC on May , I' 19, 1980, by CYAPC0 (See Exhibit 7) were discussed with M. Hills.and M.
, ! Pitek, NUSCO representatives. The investigator was pravided a data sheet (See Exhibit 12) showing the calculations for the shipment on May 1, 1980 } of bundle number H07 and also two later shipments. The decay heat content ' values are shown using the 1971 and 1979 ANSI 5.1 data. The revised 1971 data are similar to the revised 1973 data as stated in the letter dated May 21, 1980. stated that the calculations were performed after Battelle notified them of their calculations after receiving the shipment.
M also stated that he recalled contacting L. Danese, NAC representa-
tive, after Danese had left the Haddam Neck Plant site regarding his ' concerns related to the temperatures of the fuel and the cask.
The concerns were due to the cask being dry on the inside and the decontamination of the cask having to be repeated.
Since it was opined that the temperatures could affect the planned studies of the fuel, NAC was requested to perform calculations for the fuel and cask temperatures.. Pitek also recalled-discussions with Danese regarding the radiation levels 'on the container and discussions regarding th-possibility of having a bundle other than , l number H07. There may also have been discussions regarding the decay heat content but Pitek could not specifically recall all matters that were discussed.
The limit for decay heat generation specified by the Certificate of j Compliance is 2.5 Kw.
(See Exhibit 10).
Activities Involving Cask Model No. NFS-4 (NAC-lE) at Battelle Columbus
Laboratories, West Jefferson, Ohio The shipment of the NAC-lE container arrived at Battelle on May 2, 1980 and was p aced in tne fuel pool shortly after its arrival. There was a release of l !- radioactive material when the container was opened in the pool which was reported to the NRC, Region III office located in Glen Ellyn, Illinois on June - 27, 1980.
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The reported incident including the decontamination of the NAC-lE container was examined by Region III and is described in Inspection Report No. 70-008/80-02.
Activities Involving Cask Model No. NFS-4, Serial No. NAC-lE at Oyster Creek The NAC-lE container, described as empty on the bill of lading (See Attachment i, Exhibit 8), arrived at the Oyster Creek Nuclear Generating Station operated by Jersey Central Power and Light Company (JCP&L) on July 23, 1980. A survey
by the licensee found that one smear of the container was 23,000 dpm/100 cm
which exceeded the 22,000 dpm/100 cm NRC reporting level and DOT transportation limit (DOT 49 CFR 173.397(b)) for an exclusive use shipment. The survey results were reported to the NRC (See Exhibit 9).
A radiation specialist from NRC, Region I, was onsite at Oyster Creek when the container arrived. The inspector surveyed the container shipment and found a dose rate measurement of 240 mR/hr in one small area underneath the trailer.
The limit specified in DOT 49 CFR 173.393(J)(2) is 200 mR/hr.
The handling of the container by the licensee was examined and information was also obtained on the container departure from Oyster Creek.
The informat'on obtained by the , radiation specialist was provided to the investigator in a memorandum (See Exhibit 8).
As noted in Exhibit 8, the radiation specialist also contacted representatives of NAC and Battelle during the Oyster Creek Inspection. The radiation specialist also discussed the container survey results with IE:HQ while at the site. The radiation specialist also informed T. Emswiler of Battelle that the Battelle survey prior to shipment showed the maximum radiation level at the opposite end of the container than that found during the survey at Oyster Creek, which indicated that the source had moved during transport.
A copy of the information contained in Exhibit 8 was provided to Region III by memorandum dated October 28, 1980, to confirm previous telephone discussions which began on July 24, 1980, with notification regarding the container arriving at Oyster Creek in excess of DOT ani NRC reporting levels of radiation and contamination.
On October 9,1980, R. Smith, NRC Investigator was at the Oyster Creek Plant to review the circumstances surrounding the use of the container. The following , individuals were contacted: ,
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- J. Molnar, Core Manager R. Panciera, Acting Supervisor, Radiological Operations D. Turner, Manager,_ Radiation Control
- During subsequent' telephone discussions.
D. Turner provided the following information in substance: Battelle had shipped the NAC-1E container to the Oyster Creek Plant in order for JCP&L to ship a bundle of irradiated fuel to Battelle for study. This was a study being conducted by the fuel fabricator regarding cladding failure.
When the shipment arrived on July 23, 1980, it was parked in an owner controlled area which is outside the plant protected area. Shipments are surveyed at this location prior to entering the plant. When the shipment surveys on July 23 and 24,1980 revealed reportable levels of radiation and contaminat'on, the shipper (Battelle) was notified. JCP&L refused delivery of the shipment, primarily because of the radiation and contamin-ation levels, but agreed with Battelle and NAC to provide facilities for external decontamination of the cask.
The container was brought inside the plant area about July 25, 1980.
On July 29, 1980, JCP&L issued Radiation Work Permit No. 126480 (RWP) to, " Decontaminate spent fuel cask and load on trailer." The delay in beginning cask decontam t tion was due to: the discussions between JCP&L, Battelle, and NAC relative to the cask decontamination; a lock pin missing from the cask lifting yoke; and material located on the refueling floor had to be moved from the cask travel path.
The cask was removed from the transport trailer and moved to the rei'ueling floor of the reactor building for decontamination.
The trailer remained in the railroad airlock while the cask decontamination was in progress.
Survey No. 9003-80 (See Exhibit 8 - Attachment IV) shows one location on the cask with a radiation level of 2 R/hr.
Decontamination was performed by Battelle representatives.
The investigator examined the RWP and surveys related to the cask, yoke, and truck trailer.
Radiation surveys of the truck trailer and lifting yoke are shown in Exhibit 13.
The transport trailer was decontaminated
to 1000 dpm/100 cm or less and the lifting yoke was placed in a sealed container to be transported with the cask.
Contamination levels on the
outside of the sealed container were less than 1000 dpm/100 cm on the departure survey.
Lead shielding was placed on one area of the trailer to reouce radiation levels underneath the trailer to 60 mR/hr.
The Bill of Lading for the NAC-lE container departing Oyster Creek on August 15, 1980 (See Exhibit 14) was described. The notation on the Bill
of Lading regarding the original shipment being refused was confirmed by Turner.
As agreed between NAC and Battelle a stop off at West Jefferson, Ohio is also noted. This was for the purpose of removing the lifting yoke for storage at the Battelle facility.
The Radiation Protection Manager at San Onofre had contacted Turner after receiving the NAC-lE container. Turner informed him that JCP&L had refused to accept the container and that it had been brought onsite for decontamination by Battelle.
He also discussed the departure surveys being observed by an NRC Inspector (See Exhibit 8, Attachment V). The Manager from San Onofre informed Turner that arrival surveys differed from departure surveys as follows: JCP&L San Onofre Cab of truck 1 mR/hr 4 mR/hr Front of cask cage 50 mR/hr 180 mR/hr Cask Smears <1000 dpm 7000 dpm J. Molnar discussed the NAC-lE container handling at Oyster Creek with the investigator. Molnar recalled that one other consideration by JCP&L in deciding not to use the container was the required dimensional inspection being due in a few days after they would have scheduled cask loading. Molnar also stated that NAC and Battelle representatives arrived at Oyster Creek folicwing JCP&L's refusal to accept the NAC-1E container.
During discussions between Battelle and NAC at the site, NAC decided to have the container transported to San Onofre.
Activities Involving Cask Model No. NFS-4, Serial No. NAC-lE at San Onofre On August 20, 1980, the NAC-lE container arrived at the San Onofre Unit 1 site located at Camp Pendleton, California and operated by the Southern California Edison Company.
Radiation surveys conducted by the licensee revealed a radiation level of 4.4 mrem /hr in the tractor sleeper. Tha 00T limit is 2.0 mrem /hr.
(49 CFR 173.393(j)(4)). The contamination levels did not exceed the NRC reporting limits or the DOT limits.
The handling and decontamination of the NAC-1E container at San Onofre was examined by Region V and is described in Inspection Report No. 50-206/80-26.
Activities Involving Representatives of Nuclear Assurance Corporation-(NAC) During the period from October 16 to 31,1980, two investigators from the NRC, Region II Office contacted the following individuals at the NAC office location in Atlanta, Georgia.
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R. Bonnett, Cask Technician F. Danese, Supervisor, Cask Operations C. Hoffman, Supervisor, Cask Operations J. Viebrock, Manager, Operations and Engineering The investigators were provided the following information in substance.
The NAC-1E cask had been provided to CYAPC0 by NAC and was delivered to the Haddam Neck site on April 18, 1980, by Tri-State Motor Transit. The cask had previously been used to transport a fuel bundle from San Onofre to Morris, Illinois.
Danese and Bonnett confirmed that they performed the quarterly inspection at Haddam Neck on the NAC-lE cask on April 25, 1980.
Detailed inspection disclosed that leakage of one drain ball valve was caused by a bent valve flange.
Repair could not be made with the equipment and parts immediately available.
Since the otF r drain and vent valves functioned properly the faulty ball valve was removed and replaced with a pipe plug.
Danese was present during the cask loading on April 26, 1980. The cask cavity contained a basket and spacers in the arrangement used for the shipment from San Onofre and did not require changing.
Danese and Bonnett departed from the site after the final button-up of the cask and could not recall any mention by monitoring personnel of radiation levels exceeding 10 mR/hr.
Danese was contacted by CYAPC0 on April 27, 1980, regarding the radiation levels on the cask and opined that they were not consistent with previous experience with similar fuel shipments.
Discussions continued regarding the levels, cask location, and corrective action. Two options were discussed. One involved placing additional shielding external to the cask and the other involved returning the cask to the pool and rearranging the spacers in the cask cavity.
The option of supplemental shielding was utilized.
Danese attended a meeting at Battelle on May 28, 1980, regarding the cask contamination and returned to Battelle the first part of July 1980, while the cask was being decontaminated. Danese also recalled being at Cyster Creek and was present during an NRC survey of the cask on July 24, 1980.
Following decontamination of the cask and trailer at Oyster Creek the , cask departed the site for San Onofre. A stop was scheduled at Battelle to unload a yoke and special spacer basket, but only the yoke was unloaded ' at Battelle.
The pipe plug that was installed at CYAPC0 to replace the faulty ball valve was not removed until the cask arrived at San Onofre.
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Exposure of Transport Drivers The investigator contacted the Tri-State Motor Transit Company, Joplin, Missouri, to determine the radiation exposure of the drivers involved with transporting the NAC-1E cask. All transporting of the cask from CYAPC0 until delivered at San Onofre was performed by Tri-State.
- The drivers assigned for the transport of radioactive material are issued film badge dosimeters for a 30 day period. The driver that transported the empty cask from Battelle to Oyster Creek was not badged; however, the surveys at departure and arrival indicated less than 2 mrem /hr in the truck cab.
The film badge dosimeter results indicated that the other drivers did not exceed the allowable quarterly exposure for radiation workers.
Status of Investigation This investigation is being submitted in a CLOSED status.
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EXHIBITj 1.
Letter from CYAPC0 to NRC dated February 22, 1980 (Attachment contains 10 CFR 2.790(d) information and is not includec).
- 2.
Memorandum of Telephone Discussion between CYAPCO and NRC dated April 11, 1980.
- 3.
Letter from NAC to CYAPC0 dated March 25, 1980.
- 4.
Letter from NRC to CYAPC0 dated March 31, 1980.
5.
Letter from CYAPC0 to NRC dated April 15, 1980.
6.
Letter from CYAPC0 to NRC dated April 18, 1980.
7.
Letter from CYAPC0 to NRC dated May 21, 1980.
8.
Memorandum from K. Plumlee to R. Smith dated September 17, 1981, of Container Inspection at Oyster Creek.
9.
Preliminary Notification No. 1-80-109 dated July 24, 1980, 10. Certificate of Compliance No. 6698, Revision 9.
11. Bill of Lading for Shipment from CYAPC0 to Battelle.
- 12. CYAPC0 decay heat calculations data sheet dated May 30, 1980.
i3. Radiation surveys of spent fuel cask truck and cask yoke dated August 5 and 6, 1980.
14. Bill of Lading for Shipment from Oyster Creek to San Onofre.
- Copies of these Exhibits were provided by the Connecticut Yankee Atomic Power Company.
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C C) N N E C T I C U T , -1 P O e*0 A 270 M Amif omo comme CieCf 'T 06:0i ......... r o s......ii February 22, 1980 ' Docket No. 50-23 3 . . Director Office of Nuclear Materials Safety and Safeguards , , U. S. Nuclear Regulatory Commission
Washington, D. C.
20555 Centlemen: [ Haddam Neck Plant Advance Route Approval - 10CFR73, Section 73.37(a)(1) Applicant Licensee: Connecticut Yankee Atomic Power Company, FTOL DPR-61. Northeast Utilities Service Company (NUSCO), on behalf of Connecticut Yankee Atomic Power Company (CYAPCO), is presently negotiating a contract with the Electric Power Research Institute (EPRI) and Eattelle Columbus Laboratories (BCL). under which BCL vill perform both destructive and non-destructive examinations on two spent fuel assemblies from the Haddam Neck Plant.
These examinations are in support of an e1 Torr, to cetermine the cause of fuel failures that occurred during Haddam Neck's eighth cycia of operation.
' . Successful completion of this examination progra= vill provide further bases for the achievement of high levels of fuel rod cladding integrity at the Haddam Neck Plant.
This vill minimize fission product release to the primary system, thereby preventing large man-rem exposures during refueling and maintenance operations.
Ecamination of the fuel vill provide additional data in support Of the use of similarly processed fuel pellets to other reactor fuel types, ! ~ including those clad with zircaloy.
I The Department of Energy (DOE), through Battelle's Pacific Northwest Laboratories, l has an interest in a third assembly from the Haddam Neck Plant discharged after an earlier cycle of operation.
They are investigating the effects of long-term storage of irradiated fuel, probably in anticipation of the design, construction, and operation of an away-from-reactor (AFP) spent fuel storage Tscility.
Tnis , assembly would not be returned to the 'Haddam Neck Plant.
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-2-EXHIBfT 1 ' ' Page 2 of 2 - . . To support the abuse :.. r,tioned efforts, CYAPCO it, kreby applying for route approval for the rt.ored-trip shipment of up to three irradiated fuel annemblies between the Ev3du:: I;ech Plant and bCL in Columbus, Ohio.
The proposed routing, enclosed as Attach::,ent 1, was prepared for hTSCO by Mr. T. R. E=svi? er of BCL, in accordance with hthm-0561, Appendix 2-A to comply with the requireraents of 10CFR73 It should be noted that DOE may assume ownership of one or more of the assemblies in which case it (they) would not be returned to Haddan fieck.
Confirming conversations with the Staff, CYAPCO respectfully requests that this application be given expedited review in order that these a semblies can be shipped to BCL before the planned May 2,1980 shutdovn for Haddan Neck's n, ext refueling.
The coordination of cask lease, transportation, support equipment To fabrication, and manpower cannot be finalized without this liRC approval.
facilitate this effort, approval is requested no later than March 17, 1980.
Any improve.,ent on this date would be sincerely appreciated.
If you have any questions or comments on this application, please contact us.
Very truly yours, col 3ECTICUT YAEKEE ATOMIC POWER COMPANY
W. GV Counsil Vice President Attachment By: W.
F. Fee Vice President . b
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l . '. ' EXHIBIT 2 Page 1 of 6
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g UNITED STATES p, NUCLEAR REGULATORY COMMISSION. , r,
wAsmuaTon, p. c. aosss a., a g-p ..... April 11, 1980 R ECElyED . Docket No. 50-213 APR2gIgm VICE PRE 3f0ENT LICENSEE: ConnecticutYankeeAtomicPowerCompany(CYAPCO) m Mmms & Operaties FACILITY: Haddam Neck Plant i SUBJECT: SumARY OF MARCH 27, 1980 PHONE CONVERSATION REGARDING THE MOVEMENT OF A SPENT FUEL CASK INTO THE SPENT FUEL P0OL A conference telephone call was held on March 27, 1980, between members of the NRC headquarters staff and the licensee's staff to discuss movement of a spant fuel cask into the Haddam Neck spent fuel pool. A list of the participants is included as Attachment 1.
On March 21, 1978, the licensee submitted a proposal to delete from the Haddam.
Neck Technical Specifications) the prohibition of spent fuel cas(k movecent over his spent fuel pool.
1976, in conjunctOn with a change which authorized an increase in the fu storage capability of the spent fuel pool. The staff had not acted on this However,pending completion of generic task A-36, " Control of Heavy Loads".
request the licensee recently informed us that he would like to ship several failed fuel elements from Haddam Neck to Battelle Laboratory for analysis in i April 1980 before his next refueling outage, and therefore requires evaluation of their March 1978 proposal in order to bring a spent fuel cask into the spent
fuel handling building and lower it into the pool.
The task group which is considering the control of heavy loads as a generic issue was given this request to review for technical acceptability. This group studied the licensee request and identified several areas in which they needed further information to complete the evaluation.
The licensee was supplied with a list os the NRC staff's questions (Attachment 2) prior to the conference call.
All items in Attachment 2 were discussed, and the licensee agreed to provide additional details as listed below.
The licensee will provide copies of. analyses'done to verify the structural integrity of the fuel pool in the event of a cask drop. The licensee
! stated that analyses done for plants with fuel pools of similar design showed that the pool would net be significantly damaged by the dropping of a 100 ton cask.
The NRC staff would like to know the basis and assumptions made by these other analyses and a comparison to the features at Haddam Neck.
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. -- . _ v. -,; . - -- .. .-- ' i i EXHIBIT 2 ~ , s, - 'l ' '~ ' l ~ ' "1 Page 2 of 6 '- ' . i l .. - ' ' ' s - - s s , . ' -2-
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The licensee will'iJse an NAC-1 cast. H9 will determine if the yoke is * redundant and single failure proof and will investigate the availability o,f a single failure pryaf yoke. j.,, ~ . w . ,
3.
The licensee gave a value,of 2800 lb~s. as the rating of the fuel cask cover . ~,, lifting spiderc.L He will verify thab this rating is the design rating and l not the ultimate. rating, j_ , t , ..< , The 11cesee will send the NRC a detail;ed drawing of the fuel pool area, l , ' 4.
s ! includi'ng th's; locations of all spent fuel asseb.blies and seismic restraints.
l s ' , Thedrawingwillalso' diagram.Jheloadpathofthecask,andtheorientation of the ccd trunion as it traverses the load path.
- ' , ._s 5.
The licensee will verify and document to'the NRC staff that procedures satisfyir.g staff criteria will be used for the movement of the cask and ,, will verify ar.d document the training' of the crane operators in accordance ' ' with ANSI S3D.2-1976.
's ' ' ' c ._ 6.' The licensee will verify that the cask handling yoke complies with ANSI N14.6-1978 and will send us a copy of the detailed analysis for the ' ' l yoke.
c' ' ., 7.
The licensee will verify and confirm that the crane used for handling of j the cask has been maintained and inspected IAW ANSI B30.2-1976, and thct oositive motion stops or interlocks are installed to prevent improper movement of the crane used for handling the flocr hatch cover.
8.
The licensee has not completed his evaluation of the design of the crane in comperison with ANSI 830.2-1976. When this is complete, he will for-ward it to us for review.
x 9.
The licensee will ify and document that there will be no fuel elements in the pool with U concentrations greater than 4 w/o. This is to ensure that an inadvertant criticality will not occur due to crushing and a change in fuel geometry if the cask were to drop.
The licensee indicated that he would provide the requested information quickly in order to allow the staff to complete its review.
E .e Ralph aruso Operating Reactors Branch #2 Division of Operating Reactors Attachments: As stated cc w/ attachments: See next page . l ...
EXHIBIT 2 . .. . - Page 3 of 6 3- - . . $ cc w/ attachments: Dey, Berry & Howard Counselors at Law U. S. Environmental Protection One Constitution Plaza Agency Region I Office Hartford, Connecticut 06103 ATTN: EIS COORDINATOR Superintendent JFK Federal Building Haddam Neck Plant Boston, Massachusetts 02203 RFD #1 Post Office Sox 127E East Hampton, Connecticut 06424 Mr. James P. Himmelwright Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 Russell Library 119 Broad Street Middletown, Connecticut 06457 Board of Selectmen Town Hall Haddam, Connecticut 06103 Connecticut Energy Agency ATTN: Assistant Director Research and Policy Development Department of Planning and Energy Policy 20 Grand Street Hartford, Connecticut 06106 Directo/, Technical Assessment , t Division Office of Radiation Programs ( AW-459) U. S. Environmental Protection Agency Crystal Mall #2 Arlington, Virginia 20460 l l l l
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. _ _ _. .. _ - _ __ _. - _ -. . - _. _ _ _ _ . . EXHIB5T 2 ~ ' '
' P.a ge 4 o f 6 , ATTACHPINT 1 LIST OF ATTENDEES . NRC
' - e R.-Caruso H. George , H. Shaw F. Clemenson J I [ CYAPCO ' T. Murray l R. Eppinger i 3. Radder il A. Puri i
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! - - ~ ' ~ ~~~ ~~~EXMIBIT 2 l Page 5 of 6 - - l ATTACHMENT 2 HADDAM NECK - CASK HANDLING OPERATIONS 1.
Indicatewhetherafailed-fuelcontainerwillbeusedformovementofthh damaged fuel.
If it will be, describe the path to be followed and extent l to.which staff positions 1 through 5 of Enclosure 2 will be satisfied for movement of this container, i t 2.
Identify the model cask that will be used for shipment of the fuel.
3.
a) Identify the weight of the. hatch cover that is moved to the roof to allow handling of the cask.
l b) Identify where this load is store'd on the roof.
l 4.
a) Identify the weight of the spent fuel cask cover.
b) Identify the crane used for handling this cover, and the defined safe load path for its movement.
5.
Identify what safety-related equipment (including cabling) is located in the area below the location where the spent fuel cask is loaded onto the - , transfer buggy.
6.
The response to Question 1 contained in the May 14, 1974 letter from Connecticut Yankee makes reference to analyses docketed for other plants.
Describe the assumptions and approach used for the reference analysis and by whom that analysis was made. Describe the similarity of the assumptions made for that analysis and the working condicions in the Haddam Neck plant.
7.
Verify that procedures are developed and followed for the proper handling of the spent fuel cask and related heavy loads (such as the hatch cover or the spent fuel cask cover), and that these procedures include: identi-fication of proper equipment and components for performing these operations; , l required inspections before movement of the load and related acceptance i criteria; the steps and proper sequence to be followed in handling the load; definition of the safe load path; and special precautions.
8.
Verify that operators that will handle the cask and related heavy loads are trained and qualified, prior'to handling these loads, and conduct themselves in accordance with Chapter 2-3 of ANSI B30.2-1976, " Overhead and Gantry Cranes".
, 9.
Verify that the yoke used to handle the cask satisfied the guidelines of ANSI N14.6-1978; however, the stress design factor stated in Section 3.2.1.1 of ANSI N14.6 should be based on the combined maximum static and dynamic loads that could be imparted on the handling device based on the character-istics of the crane which will be used.
10. Verify that the slings er handling devices used for movement of the hatch cover and spent fuel cask cover (if different from the cask yoke) are installed and used in accordance with ANSI B30.9-1971, " Slings".
~ EXHIBIT 2 .,,.. Page 6 of 6 - . . 2- - 11. Verify that the crane (s) used for handling of the spent fuel shipping
, cask and related heavy loads are inspected, tested, and maintained in accordance with Chapter 2-2 of ANSI B30.2-1976, with the exception that tests and inspections should be performed prior to use where frequency of crane use for these loads is less than the specified inspection and test frequency.
12. Verify that.the crane design satisfies the guidance of ANSI B30.2-1976, Chapter 2-1.
Provide justification for those provisions that are not met.
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Page 1 c? 2 . , EXHIB?T 3 Nucler Assurr a Cwomatim 24 Executive Park West Attor, a. Georgia 30339 (404)325-4200 Telex: 549567. 542703
715 F onzon Dme Grand.Junc: son. Colorada 81501 (303)245-4320 TWX 910929G334 Weinbergstrasse 9 8001 Zunch. Switzerland ' (01)470844 Teler 57275 March 25, 1980 FLD/80/15/ETS M Northeast Utilities Services Lompany P. O. Box 270 Hartford, Connecticut 06101
Dear Mr. Pitek:
As indicated below, we have provided a suggested response to certain of the questions posed by the NRC with regard to cask handling operations.
The numbering system used corresponds to the numbers used in the telecopied request.
HADDEM NECK NRC Questions - Cask Handling Operations Requests for additional information: 1.
A failed-fuel container will not be used for movement of the damaged fuel. The NFS-4 (NAC-1) model shipping cask is a zero release cask in its shipping configuration.
Failed fuel shall be loaded into the cask underwater in the fuel pool. All movements after loading are performed with the cask. Consequently, no failed-fuel container is required.
~ 2.
The cask model number is NFS-4 (NAC-1), licensed under Certificate of Compliance Number 6698, Revision 9, dated December 12, 1979.
4.
(a) The estimated weight of the spent fuel cask cover is 750 pounds.
(b) Typically, the auxiliary (5-ton) crane is used for handling the i cask cover.
Staff Positions 1.
Cask Handling and Loading Procedures have been provided to plant person-nel. These procedures identify necessi y equipment to adequately handle the cask. Selected aspects of *ne procedures can and should be incorporated into plant operations procedures and be reviewed by the Facility Review Group.
~ ~ M EXHIBIT 3 - - - - Page Two Page 2 of 2 March 25, 1980 . 2.
Nuclear Assurance Corporation can provide qualified cask har.dling assis-tance to the plant.
3.
The yoke used to handle the cask satisfies the guidelines of ANSI H14.6-1978 and specifically meets the (static load) requirements of Section 3.2.1.1.
However, we note that the referenced Section (Section 3.2.1.1.)
has no requirement that stress design factors include static and dynamic loads. Dynamic loads are a function of crane characteristics.
4.
A lid lifting spider is used to handle the cask lid. The lifting spider is attached by four 1" bolts, is load-tested to 2000 lbs. and is con-structed of C 1020 steel.
Please let us know if additional information is required.
Sincerely, , NUCLEAR ASSURANCE CORPORATION m o,, b um Larry Danese Supervisor, Cask Operations FLD:cnr -
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- EXHIBIT 4 Page 1 of 2 ' ,,, s u c..,t.
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L .- " f wucLEAR REGUI. ATONY CO.U.USSION ,S y,3,f),. i . w.:.ws:am. o. c. mx .:,g (..*. > / e; . . ./ . IJAR 3 j 17cu- ' RECElVED
. I SCPL:CPJi . 50-213
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h . ' Connecticut Yankee Atoaiic Power Company !!udur Enginc:iint e OiEMIICDS I ATTih W. G. Counsil ~ - Vice President ,, ' P.O. Box 270 - Ilartford, Connecticut 05101 s . - . Gentiemen: , This is in regard to your request for ap;,roval of a route to be used for transport of spent -recctor fuel n', cor.tcir.::.. in ycur isnc cf Teb: cary 22.
1980 Subject: Haddcm iteck Plaw: - Advr.r v_d nrute A;pr cal - 20 CFF. 73, Section 7's.37(a)(1) - Applicant Licensee: Cenn:.cticut Yar ;.ee,Atcaic. Puwcr Com,any, FTOl. DPR-61.
' The " Proposed Pouting - Routing Plan" sub:aitted in the at'.:.ch:2rt to ycur February 22,19E0 lettcr is judged to tect the regulatory requirements in accordance with 10 CFR Part 73.37 and accordingly is epprcycd.
Please note that assuring highway safety is tke' responsibility of the licensee and carrier and an approval is not int:.nded to provide relief , in this regard.
Furthermore, thc approval decs not guarantee that there will be no local or state legislation applicabic to the route that - restricts or prohibits the movement of radioactive material., During the spring months when incicn2nt weather with accr".pcnying hazardous road cc::ditions can occur Gith short notice, the apprcpricte state police , should be contacted with regard to road conditions before a shipment ! c commences.
. The ' initial arrangements with law enforcement agencies clong the route,. l as required by 10 CFR Part 73.37(a)(2), have been completed by the !!RC staff.
Data relating to these arrenec ents and a copy of the approved route are enclosed.
This infor: nation is to be incorpera:cd into your ! shipment plan and provided to your carrier clong with instructions . regarding its use, l, Picase note that the notification requirements of 10 CFR Part 73.72 for r cach i.idividual shipment still apply.
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, g.. EXHIBIT 4 l . ' ~ Page 2 of 2 { , -
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. - , 22, 1980 contcins inforriation Since the attachr.ent to your letter of february of a type specified in 10 CFC Part 2.790(d), it is deemed to be cor.mercial or financial inforx, tion within the meaning of 30 CFR Part 9.5(a)(4) and steall be subject to disclosure only in accordance with the provisions of - 30 CTR Part 9.12.
For the same reason, the enclosure to this letter is , being withheid.
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Sincerely, - ' - . . , , < f $ Geor. e b'. McCorkle, Chi 6f f Phy ical Security Licenhing Branch - l ' - . Division of Safeguards, NMSS - - e , .
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i EXHIBIT 5 Page 1 of 1 C O N N LE C T I C U T Y AN K EE AT O MIC POWER C O N' P A N Y $'l{g@V - ~ B E R t.I N. C O N N E CTIC U T % h p_ o. som 2To H antrono. CONNECTICUT ostOI m...... 2o3-666-6911 April 15, 1980 Docket No. 50-213 Mr. Ioyce H. Grier, Director - Region I - Office of Inspection and Enforcement U. S. Nuclear Regulatory Ccamission 631 Park Avenue King of Prussia, PA 19h06 Gentlemen: .. Haddam Neck Plant FTOL No. DPR-61 Intent to Ship Special Nuclear Material In accordance with the requirements of 10CFR73.72,' Connecticut Yankee Atomic Power Company (CYAPCO) is hereby providing notification of our intent to ship special nuclerr material in the form of three (3) irradiated fuel assemblies.
Each of these assemblies vill be shipped by truck, owned by Tri-State Motor Transit Company of Joplin, Missouri, in three separate shipments from CYAPCO's Haddam Neck Plant in Haddam Neck, Connecticut, to Battelle Celunbus Laboratories in Columbus, Ohio.
Shipments are scheduled for departure from Haddam Neck at 8:00 a.m. on April 22, 28, and May 2,1980, with scheduled arrival at Battelle Colu= bus 22 hours after departure in each case.
. Return shipments of two (2) of the assemblies are expected, but have not yet been scheduled.
CYAPCO intends to send notification of those shipments when they are scheduled.
. .., The ReS on 1 office was notified by phone today, April 15, 1980.
i Very truly yours, CONNECTICUT YAN9.EE ATOMIC POWER COMPANY
W. G.(founsil Vice President - / 2.Q - - a ay , W. F. Fee Vice President , e .. - -.... . - s
. ~ - EXHIBIT 6 Page 1 of 17 CONNECTICUT YA N K EE ATO MIC POWER COMPANY BERLIN. CON N ECTIC UT P.O.som aTo N ARTroRO, CONNECTICUT 08 tot 6.
w e... 203-488 88I1 April 18, 1980 Dociet No. 50-213 Director of Nuclear Reactor Pagulation Attn: Mr. D. L. Zi m nn, Chief Operating Reactors Branch H2 U. S. Nuclear Regulatory Co= mission Washington, D.C.
20555 Reference: (1) D. C. Switzer letter to D. L. Zie= ann dated March 21, 1978.
Gentle =en: Haddam Neck Plant Additional Information in Support of Proposed Chnnges to Technical Specifications In Reference (1), Connecticut Yankee Atomic Power Company (CYAPCO) forwarded proposed Specifications to delete the prohibition of spent fuel cask movement over the spent fuel pool in Technical Specification 3.13H and substitute additional res',rictions regarding their movement. Also included were the results of the cask drop analysis.
I In late February,1980, CYAPCO requested that the NRC Staff expedite their re-view of the propos.ed changes in order that shipment to Battelle Columbus of three (3) spent fuel assemblies from previous cycles at Haddam Neck could be acco=plished before the upcoming May 3,19?0 shutdown for refueling.
At that time, it was emphasized that the shipment of these assemblics and the work to be done on them is of interest, not only to CYAPCO, but also to the Electric Power Research Institute (EPRI), the U.S. Department of Energy (DOE), and the industry in general.
Two of the assemblies are being shipped as part of the joint CYAPCO-EPRI fuel examination program directed toward definitively detemining the cause of Batch 8 fuel failures experienced in 1979 Shipping and srnsequent post-l irradiation examination are critical to assuring against future failures and minimizing radioactive contamination. The third assembly is a thoroughly pre-characterized assembly from Cycle 2 operation at Haddam Neck, which bac been A in the spent fuel pool for approximately seven years. DOE is interested in investigating the effects of long-tem storage on spent fuel in connection with the Away-From-Reactor (AFR) spent fuel storage program.
-_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ EXHIBIT 6 - - .
Page 2 of 17-2- . On March 27, 1980, a conference telephone call was held to discuss questions from the Staff which had been informally transmitted to CYAPCO/NUSCO a week earlier. Further questions and clarification requests resulted from that call. Answers have been informally transmitted for your review previously and are provided formally as Attachnent 1 to this letter.
On April 3,1980, CYAPCO advised the Staff by telephone that we were revising the request for a permanent Technical Specification change to a one-time approval, vaiver, license condition, or whatever would be the most expeditious method of modifying the Technical Specifications to allow movement of the spent fuel .ask and accomplishment of the chipnents. This revision was the result of addi-tional concerns, raised internally at NUSCO, regarding potential criticality in the event of a cask drop.
At this time, CYAPCO respectfully reiterates its requests for expeditious Staff action on this topic. To avoid a very long delay, shipment of the three fuel assemblies must be accomplished before the scheduled refueling outage.
Yeu: continued cooperation is sincerely appreciated.
Very truly yours, CONNECTICUT YANIGE ATOMIC POWER COMPANY Y _. $g / .l*
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/fd ' . a W. G.-Couns'il ' Vice President Attachment . . - _ - _ _. - - - _ _ _ _ _ _ _ - _ _ _ _. - _ _ _. - _ _ _ _ _ _ - _. - _. _ _ _. _ _. _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _. _ _ _. _ _ _ _.. _ _. _
N l- ' - - Page 3 of 17 - , DOCKET NO. 50-213- > , I . I I i .J.
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ATTACIDfENT 1 , NADDAM NECK PLANT ' i . ADDITIONAL INFORMATION IN SUPPORT OF PROPOSED CHANGES I i ! 'IO TECHNICAL SPECIFICATIONS i t !, s i , t i ! ! , , i i l l l . l l l; L l . !
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EXHIBIT 6 .. . /STACHMENT 1 Page 4 of 17 QUESTIONS TELECOPIED FROM NRC AS " ENCLOSURE 1", HADDAM NICK, CASK HANDLING OPERATIONS, REQUEST FOR ADDITIONAL INFORMATION Question (11 Indicate whether a failed-fuel container vill be used for movement of the da= aged fuel.
If it will be, describe the path to be followed and extent to which Staff Positions 1 through 5 of Enclosure 2 vill be satisfied for movement of this container.
Response CYAPC0 vill be using zero leak cask, and so, will not need a failed-fuel container.
i i Question (2) Identify the model cask that will be used for shipment of the fuel.
l Response CYAPCO vill use an NAC-1 cask, licensed as NSF h.
Question (3) a) Identify the vei ht of the hatch cover that is moved to the roof to
allow handling of the cask.
b) Identify where this load is stored on the roof.
l Response There is no hatch cover which is moved to the roof. There is a hatch in the roof whose cover slides, on tracks, to the side of the opening out of the way.
This is at elevation 75'6".
There is also a hatch in the ficor at elevation h7'0", whose cover is lifted off and set aside on the same floor, away from the l pool and out of the way.
l l Question (h) a) Identify the weight of the spent fuel cask cover.
b) Identify the crane used for handlin6..his cover, and the defined safe load path for its movement.
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-2-EXHIBIT 6 , . Page 5 of 17 , Response The spent fuel cask cover veighs 750 pounds. It is removed.from the cask using the Refuel Building's 3-1/2 ton crane. The defined path is indicated on Figure 2.
Question (5) Identify what safety-related equipment (including cabling) is located in the area below the location where the spent fuel cask is loaded onto the transfer buggy.
Response There is none.
Question (6) The response to Question 1 contained in the May 14,197k letter from Connecticut Describe the Yankee makes reference to analyses decketed for other plants.
assumptions and approach used for the reference analysis and by whom that analysis was made. Describe the sLnilarity of the assumptions me.de for that analysis and the working conditions in the Haddam Neck Plant.
Response The referenced analysis was done by Stone and Webster Engineering Corporation for a spent fuel pool of similar design. The assumptions and conservatisms made in that analysis were as follows: a) All the kinetic energy of the cask at impact is absorbed by crushing of the concrete in the pool floor only -- the cask was assumed to be infinitely stiff, not absorbing any kinetic energy itself, b) The concrete strength used (3000 psi) is the design strength, that is, the strength after 28 days. The actual strength of the concrete increases , l over time.
c) Although the cask drop is a dynamic event, the static crushing strength of the concrete was used.
d) The line of impact goes through the cask's center of gravity and the cask's edge which penetrates into the concrete slab for the duration i of the event.
e) The crush zone is a L5' cone with a bottom diameter of 15 feet (radius f at rock contact = 7 5 feet).
l l . . .
EXHIBIT 6 ' ' ' Page 6 of 17-3-f) The concrete in the crush zone is assumed infinitely permeable, g) The flow in the rock is calculated using hemispherical artesian flov equation.
The resulting leak rate calculated was approximately 0.k gpm.
The referenced plant's spent fuel pool is essentially identical to that at Haddam !{eck.
Both are poured directly on bedrock witn e'. minimum concrete thickness of six (6) feet on the pool floor.
It should be noted that the referenced analysis assumed a 100 ton cask is dropped from a height of 42 feet (k feet through air, 38 feet through water). The cask to be used st Hadda: Neck veighs 25 tons and could be dropped from a height of only approxi=ately 38 feet (h feet through air, 33-1/2 feet through water). The concrete used in the Haddam Neck pool has a design strength of 3000 psi, and has been in place since 1967 -
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- - - - - . 4- , , EXHIBIT 6 QUESTIONS TELECOPIED FROM NRC AS "ENCICSURE 2", HADDAM NECK. CASK HANDLING OPERATIONS. REQUEST FOR ADDITIONAL INFORMATION i . Question (1) Verify that procedures are developed and followed for the proper handling of the spent fuel cask and related heavy loads (such as the hatch cover),.and that these procedures include: identification of proper equipment and . components for performing these operations; required inspections before movement of the load and related acceptance criteria; the steps and proper ' sequence to be followed in handling the load; definition of the safe load path; and special precautions.
, Response Procedures for cask handling operations were supplied by NAC. These have been incorporated into the site-specific procedures for the Haddam Neck The path of the cask is indicated on Figures 1 and 2 of this Plant.
Attachment. The cask cover will be stored in a position, suspended off the . fuel crane bridge, which will be above the seismic supports, not above the , fuel.
The yoke will be oriented (by procedure control) in the East-West direction within the Spent Fuel Building. Thus, if one end of the yoke should break, when the cask is over the pool, the cask would tip eithe:- in the East direction j away from the fuel storage pool, or in the West direction, over the fuel rack seismic restraints, thus, providing additional protection against the possi-bility of the cask tipping towards the stored fuel assemblies.
The cask is restricted from motion over the fuel storage racks by the roof hatch opening area.
(See Drawing 16103-2703). As a precautionary measure, CYAPC0 can put an additional man in the crane cab with e. walkie-talkie and another man stationed inside at the crane's main power breaker. Thus, in the extremely unlikely event of the failure of the crane directional control, with coincident failure of the crane stop button and the power breater the cab, , the second man in the cab can relay a message to the man inside and he, in turn, , will open the main breaker.
Drawings, numbers 16103-27036 and -59005 (partial), are provided for information.
The partial of 16103-59005 is the fuel pool only, showing the seismic restraints.
Question (2)
Verify that operators that will handle the cask and related heavy loads are trained and qualified, prior to handling these' loads, and conduct themselves in accordance with Chapter 2-3 of ANSI B30.2-1976, " Overhead and Gantry Cranes".
- Response CYAPCG will ensure that operators are trained.and gr lified, and conduct themselves in accordance with Chapter 2-3 of ANSI B30.2-1976.
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-5-EXHIBIT 6 ' Page 8 of 17 Question (3) Verify that the yoke used to handle the cask sat isfies the, guidelines of ANSI N1h.6-1978, however, the stress design factor stated in Section 3 2.1.1 of ANSI N14.6 should be based on the combined maximum static and dynamic loads that could be imparted on the handling device based cn the characteristics of the crane which will be used.
. Fesponse The NAC-1 cask Lifting Yokes to be used at Connecticut Yankee, for the failed fuel handling operation, are Y-3 and SY-1 designation. Each of these lifting yokes was designed by Nuclear Fuel Services Company (NFSC). Their design criteria (attachment) were as indicated in Para 6raph h.2 of NAC Document 101-0-A2, using a design load of three (3) times the load cask dead weight (8 150,000 pounds) with design stress less than the minimum yield stress.
For a design load of 5W (250,000 pounds) the design stress is less than the ultimate strength of yoke material (92,600 psi).
The adequacy of this design was demonstrated by load test after fabrication (to 150 percent of Icaded cask weight) per ANSI 1h.6 guidelines.
As indicated in Section k.2.h of the design criteria, minimum yield strength of 50,000 psi was utilized. Material certification performed by United States Steel (USS) indicated the minimum yield strength to be 61,700 psi. Utilizing the higher yield strength induces an additional safety factor of 23.4 percent.
over the safety factors established using a 3W design load and a 50,000 psi yield strength.
NAC's recertification of the design modifications performed on the origital NFS yoke designs rescits in a minimum safety factor of 1.k for the yield stress criteria using the 3W design load.
(NFS's original design basis was not modified.)
t Based on the above discussion, it is NUSCO's opinion that the NAC yokes adequately meet ANSI 1h.6 requirements.
In sddition, the yokes will have a minimum safety factor of 1.72 (1.h * 1.23) for a 3W design lead. The minimum dynamic load factor, without exceeding the yield stress criteria, for a 1W load is 5 0.
The cask does not have a redundant or single-failure-proof yoke. There is none available in the industry.
Question (4) Verify that the slings or handling devices used for movement of the hatch cover and spent fuel cask cover (if different from the cask yoke) are installed and used in accordance with ANSI B30 9-1971," Slings".
- .- ,, . -, -, -,, -. , _ +.-. -, ~_..., - -- , - - - - - -.. - - - - -
EXHIBIT 6 - . Page 9 of 17-6-Response . CYAPCO vill ensure that the alings used for movement of the cask cover are used in accordance with ANSI B30 9-1971. The lifting spider has a design rating of 2,800 lbs., is lead tested to 2,800 lbs., and has a factor c~f safety of 3.
Question (5) Verify that the crane (s) used for handling of the spent fuel shipping cask and related heavy loads are inspected, tested, and maintained in accordance with Chapter 2-2 of ANSI B30.2-1976, with the exception that tests and inspections should be performed prior to use where frequency of crane use for these loads is less than the specified inspection and test frequency.
Response The fuel building auxiliary hoist and the large overhead crane, which vill be used for handling heavy loads and the spent fuel shipping cask, were recently (April 11, 1980 and April 15,1980) inspected by Dwight-Foote, Incorporated, of Berlin, Connecticut. Their representatives have indicated that inspection requirements of ANSI B30.2-1976 were met or exceeded, and that certification to that effect vill be provided. It is CIAPCO's intent to. load-test the crane and complete lifting rig to approximately 150% of the shipping cask / fuel assenbly weight before cask handling operatioAs begin.
Question (6) Verify that the crane design satisfies the guidance of ANSI B30.2-1976, , Chapter 2-1.
Provide justification for those provisions that ara not met.
Response ! The crane was purchased in 1964 It was designed to the following specifica- ' tions applicable at that time: a) AISC Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings.
b) Specification for Steel for Bridges of the ASTM, Serial Designation A-7.
c) Electric Overhead Crane Institute Specifications.
d) General Information for Standard Industrial Service Electric Overhead Traveling Cranes.
,
- e) ASA Standards.
l .
EXHIBIT 6 ' ' Page 10 of 17 ' -7_ As indicated above (Question (5)), inspection results meet 'or exceed the requirements of ANSI B30.2-1976 and CYAPCO intends to load test the crane and lifting rig before cask operations are started.
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emwtrtanr,, Page 13 of 17 - - ' Ection: EDA Docum nt No.
101-0-A2 Page 1 Title: Design Analysis for NAC-1 C:sk Lifting Yoke Modification Revision No.
Date Prepared By: C. C. Hoffman
12/10/76
. 1.0 PURPOSE Provide Design Analysis of the NAC-1 Cask Lifting Yokes Y-3 and Y-4.
2.0 APPLICABILITY AND SCOPE This Analysis applies to the NAC-1 Cask Lifting Yokes Y-3 and Y h.
3.0 DEFINITIONS None 4.0 ANALYSIS 4.1 Design Bases h.1.1 The design of the NAC-1 Cask Lifting Yoke Y-3 - and Y 4 is based upon Nuclear Fuel Services, Inc., lifting yoke design, Drawing No. NSF E10099, Rev. 8.
I Five yokes built to this NFS design have been proof-load tested to 130,000 pounds and used in service with the NFS k and NAC-1 casks for several hundred shipments . without any evidence of failure.
k.1.2 The design of the NAC-1 Cr.sk Lifting Yoke Y-3 and Y 4 is identical to NSF E10399, Rev. 8, design with the exception that width of the bo'ok-eye opening is increased from 10 inches to 13 inches.
. ! k.1.3 Material specifications are identical to those utilized on the original NFS design.
l - . -..
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.
_ - - Page 14 of 17 Section: EDA Document No.
101-0-A2 Page 2 f Title: Design Analysis for NAC-1
Cask Lifting Yoke Modification Revision No.
Date Prepared By: C. C. Hoffman
12/10/76 . k.l.h A stress analysis covering only the modification incor-porated, the increase in vidth of the book-eye, is performed.
h.2 Design Criteria k.2.1 Yoke design is to the yield point (or less) of the material for the design load.
h.2.2 Static load (cask veight) is 50,000 pounds.
, k.2.3 Design load is 150,000 pounds (3 times the actual load).
h.2.4 Material is Corton-B with a minimum yield strength of 50,000 psi.
h.2 5 Ioad test is required 75,000 pounds (150% of loaded NAC-1 cask veight).
h.2.6 The yoke is a safety-related item and quality assurance control and documentation is required covering materials, fabrication, and testing.
. h.2 7 Operating environmental conditions are identical to those previously experienced in service for the NFS E10099, Rev. 8, yoter.
i _____ _ _ _ - - - - _ - - - _ _ _ _ _ _ _ _ _ _ - - _ - - - - - _ - _ - - - - - - - - - - - - - _ - - - - - - -. - - - _ - _ - - - - - - - - - - - - - - - - - - -- - J
' inM10 Dnna U - - - Page 15 of 17 Section: EDA Document No.
101-0-A2 Title: Design Analysis for NAC-1 Page 3 Cask Lifting Yoke Modification Revision No.
Date Prepared By: C. C. Hoffman
12/10/76 k. '3 Design Analysis The design of the NAC-1 Cask Lifting Yokes Y-3 and Y-4 is identical with one minor exception, to the design of the NFS-k Cask Lifting Yoke. As noted in the design bases, load t yting and service use has demonstrated the adequacy , of that design. This analysis covers only the effect of the hook-eye modification to the design.
. O
. MHIWII 6 Page 16 of 17 , . . . , . . . I.
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n - .p , EXHIBIT 7 Page 1 of 2 ' ' ~ 30.
CONNECTICUT YANKEE AT O M IC POWER COMPANY s/ .
BERLIN. CONNECTICUT ^\\ e o som 27o mantrono. countet cut ostos
,... -. o r...... n , May 21,1980 . Docket No. 50-213 B10003 . . Mr. Boyce H. Grier, Director Region 1 Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission 631 Park Avenue King of Pru'ssia, PA 19406 Ref erences: (1) W. G. Counsil letter to Director, Of fice of Nuclear Materials Safety and Safeguards dated February 22, 1980.
(2) W. G. Counsil letter to Director, Of fice of Nuclear Materials Safety and Safeguards dated March 24, 1980.
(3) W. G. Counsil letter to B. H. Grier dated April 13, 1980.
Gentlemen: , Haddam Neck Plant Information on Spent Fuel Shipment Connecticut Yankee Atomic Power Company (CYAPCO) recently shipped three - spent fuel assemblies to Battelle Columbus Laboratories in three separate shipments in an NAC-1 single assembly spent fuel shipping cask.
The NAC-1 cask was originally licensed for 11.5 Kw of residual heat with an assembly - shipped wet.
However, NRC recently imposed a restriction to 2.5 Kw when the decision was =ade to ship assemblies dry because of potential leakage concerns under certain accident conditions.
The restriction is anticipated to be temporary while the cask licensee completes the re-licensing ef fort.
The first assembly shipped from the Haddam Neck Plant was a failed assembly f rom Batch 8.
Our calculations indicated decay heat content of 2.09 Kw using the proposed 1973 ANS1 5.1 data (which is also the basis of Appendix K). CYAPCO was notified by Battelle on May 12, 1980 that a post-receipt confirmatory calculation indicated a decay heat rating of greater than 2.50 Kw, using the recently approved 1979 ANSI 5.1 standard.
We have subsequently confirmed that actual heat generation is 3.50 Kw using the , ' 1979 ANSI 5.0 standard.
Independent calculations were performed by NAC, the cask licensee, to check and confirm this figure.
- .
I .
ETRJIBllY Y
. - 2-Page 2 of 2 . , .. On establishing the accuracy of the 3.5 Kw figure for the first assembly, CYAPCO contacted NRC's Office of Nuclear Material Safety and Safeguards and Region I Headquarters, on Friday, May 16, 1980,_ to notify the respective offices that the cask's license decay heat limic had been violated.
This letter is being docketed in accordance with the provisions of 10CFR Part 71 and DPR-61.
It was noted at that time, and is emphasized at this point, that the shipment in-question was accomplished with no adverse effects on the health and safety of the public, and that no appareat damage to the cask'resulted.
State authorities were also informed.
Calculations have also been re-done for the second and third fuel assemblies . shipped, and verify that both assemblies were less than 2.5 Kw.
In addition, an investigation has been initiated by the cask licensee to determine the extent of the ef fects on the cask, if any.
Because there are no adverse consequences, associated with this incident, no further action on either CYAPCO's or the Staff's part is judged to be necessary.
Very truly yours, CONNECTICUT YANKEE ATOMIC PO'-7.R COMPANY . U/.l'1 i,! /',,/ 7TMX W. G. Counsil Vice President .
r EXHIBIT 8 Page 1lof 34 ' , 17 SEP 1981 Docket No. 50-219 MEMORANDUM FOR: R. H. Smith, Investigator FROM: Karl E. Plumlee, Radia: on Specialist, Facilities Radiological Protection Section, DETI SUBJECT: REVIEW 0F LICENSEE MANAGEMENT OF AN EMPTY SPENT FUEL SHIPPING CONTAINER, NAC-lE, RECEIVED CONTAMINATED ON July 23, 1980, AT THE OYSTER CREEK NUCLEAR POWER PLANT (DN 50-219) SUMMARY: The licensee's canagement of the NAC-lE spent fuel shipping container was observed during the period July 23 to August 1, 1930.
(Inspection Report No. 50-219/S0-26).
No noncompliance with DDT and NRC regulations was identified involving the reactor licensee.
DETAILS: 1.
Persons contacted on site in connection with the NAC-1E container.
Jersey Central Power and Licht Ccmpany (JCP&L) Ecoloyees J. Carroll, Station Manager I. Finfrock, Jr., Vice President, Generation D. Turner, Radiation Protection Supervisor General Dynamics Corporation (Electric Boat Co. Emoloyees, contracted to JCP&L) N. Gannon, Acting Group Supervisor, Field Operations (Radiation Protection) R. Panciera, Acting Field Operations Supervisor Battelle-Columbus Laboratories T. Emswiler, Senior Specialist, Nuclear Packaging and Transportation Nuclear Assurance Corocration R. Bonnett, Quality Assurance Engineer F. L. Danese, Quality Assurance Engineer
Nuclear Reculatory Commission P. J. Knapp, Chief, Radiation Support Section, FFMS Branch, Region I J. A. Thomas, Resident Inspector - - - - -
~1 EXHIBIT 8 ' Page 2 of.34 - . Memo for Mr. R. H. Sm'th
.t7 ggg 3gg, 2.
Planned Use On-Site of the NAC-lE Container The reactor licensee planned to load a spent fuel element in the NAC-lE container for shipment to Battelle-Columbus Labs (Letter dated July 17, 1980).
3.
Adherence to Receivino Procedures 2 3wj, The licensee routine receiving survey identified-a 23,000 dpm/100cm p on the front of the container collision shield.
Specifically, this was at ~ a bolt head. The container was'inside a cage on a sole use vehicle. Attach-ments I and II are copies of the shipping records and the receiving survey records.
The licensee notified the NRC resident inspector and he assisted the NRC-radiation specialist with a confirmatory survey of the container while still in the cage on the vehicle, but after the licensee had wiped down the container. Attachment III is a copy of the NRC confirmatory survey record.
The confirmatory swipe survey identified a' maximum of 145-dpm/100cm
remaining on the container, and 4,991 dpm/100cm on a nearly inaccessible area of the trailer floor.
(Subsequently after lifting the container off
the trailer the licensee found >400,000 dpm/100cm on a previously inaccess- -ible area of the trailer floor.)
The licensee representative participated with the. inspectors in' conducting a thorough survey of the container and the trailer, before the removal of the container from the trailer.
The inspector identified one small area underneath'the trailer, directly beneath-the lower drain hole of the cask, in excess of the 200 mr/hr limit-on the external surface of the car or vehicle required by 49 CFR 173.393(j)(2).
This area was physically accessible.
As an example, an individual could place his head in this area if he per-i formed maintenance work on the trailer.
' Using an NRC XETEX survey instrument, which has a small detector, this dose i rate measurement was 320 mr/hr.
! I Using the licensee Eberline R0-2A survey instrument, which has a larger-detector than the XETEX survey instrument, this dose rate measurement was 240 mr/hr.
I The inspector attributed the difference in the two measurements to indicate i that the beam of radiation was narrow.
Subsequent measurements at the lower drain hole of the container indicated there was >2R/hr at contact, recorded on page 1, item 6 of Attachment IV.
' , , i
-,.. - _ _ _ _ _ _. . . . . . . . )
EXHIBIT'8-Page 3 of 34 . . Memo for Mr. R. H. Smith 3-J 7 SE? -1981~ Review of these measurements did not identify any item of noncompliance by-JCP&L, however it appeared that the transferor's compliance with. shipping.
requirements should be reviewed.
The inspector noted that the transferor was Battelle-Columbus Labs, 505 King Avenue, Columbus, Ohio, which is.in NRC:IE Region III, and the informa-tion relating to this shipment was' transferred to NRC Region III for review on October 28,.1980.
~ 4.
Reculatory Recuirements Applicable to the Transferor (Shipper) . The following requirements appeared to apply to the transferor.
10 'CFR 71.3 Requirement for license.
No licensee subject to the regulatiens in this part shall (a) deliver any licensed materials to a carrier for transport or (b) transport licensed material except as authorized in a general license or. specific license issued by the Commission, or as exempted in this part.
10 CFR 71.5 Transportation of licensed material.
(a) No licensee shall transport any licensed material outsi, of the confines of his plant or other place of use, or deliver any li nsed material to a carrier for transport,.unless the licensee complies with the applicable requirements of the regulations appropriate to the mode of transport, of the Department of Transportation in 49 CFR Parts 170-189, and the U.S. Postal Service in the' Postal Service Manual (Domestic Mail Manual), section 124.3,. incorporated by reference,.39 CFR 111.1 (1974), insofar as such regulations relate to the packaging of byproduct, source-or special-nuclear material, marking and labeling of the packages, loading and storage of packages, placarding of the transportation vehicle, monitoring requirements-and accident reporting.
49 CFR 173.393, " General packaging'and shipping requirements." (a) Unless otherwise specified, all shipments of radioactive materials must meet all requirements of this section.
! l 49 CFR 173.393 (i) and (j) require: "(i) Except for shipments described in ' paragraph (j) of this section, all radioactive materials must be packaged in suitable packaging (shielded, if necessary) so that at any time during ~ the normal conditions incident to transportation the radiation dose rate does not exceed 200 millirem per hour at eny point on the external surface ! of the package... (j) Packages for which the radiation dose rate exceeds , the limits specified in paragraph (i) of this section, but does not exceed
. at any time during transportation any of the limits specified in paragraphs-(j) (1) through (4) of this section may be transported in a transport ' vehicle which has been consigned as exclusive use (except aircraft).
Specific instructions for maintenance of the~ exclusive use (sole use) shipment controls must be provided by the shipper to the carrier.
Such instructions must be included with the shipping paper information: (1)' ' 1,000 millirem per hour at 3 feet from the external surface of the package ! l -
i L
. -.. . _. _ . . . . _ -. - _ = EXHIBIT 8 Page 4 of 34 , , Memo for Mr.'R. H. Smith.
il 7 SEP 1961 L . j (closed transport vehicle only); (2) 200 millirem per hour at any point on !: the. external surface of the car or vehicle (closed transport vehicle only); ' (3) Ten millirem per hour at any point 1 meter (three feet) from the vertical i planes projected by the outer. lateral; surface of the car or vehicle; or if-
the load is transported in an open transport vehicle; at v point 2 meters . (six feet) from the vertical planes projected from the ater edges of the , i vehicle; (4) 2 millirem per hour in any normally occup.ed position'in the
car or vehicle except that this provision-does not apply to private motor - carriers."
t
5.
Shioning Information as of Deoarture from JCP&L,-0yster Creek ' ' JCP&L. refused to accept and use the. shipping. container.
It was hoisted to - the refuel floor of the-reactor building for. decontamination by Battelle- , Columbus. Labs representatives before release for tra" sport.off site. The . licensee discovered significant levels of beta gamma radiation at duct tape over both the upper and lower drain holes.
, . -The licensee discovered that a lock pin, used in the lifting yoke for the ~ cask, was missing. Withxthe concurrence of the cask owner's representative a substitute pin was used during-hoisting.
The survey on departure from~JCP&L was performed by JCP&L personnel and war.
witness by the NRC resident inspector. Attachment V is a copy of the , . survey record.
'- h,C jy sarl E. Plumlee ' i Radiation Specialist Attachments: As Stated L ! l . 9 1
a
1
j' , i - ... . . . . .. . . . .
EXHIBIT 8 , - Page 5 of 34 , Attachment I (12 pages) s Shipping Record for NAC 1-E Cask Shipped July 22, 1980 by Battelle-Columbus Laboratories i l I r --
_ . .. . .., . - ., - . . . ' EXHIBI
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Survey Instrument:~ type (s) O L-V N 19 - _ Give exact location either by writing, attached floor plan, or drawing on back of this form.
- . . Results of Survey (Units in c/m, rad orQrad/hD or nicm /sec.)
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' .- - "Si i-r 13 M t/ , ~ . Ul ~Tl\\\\$ WG.rq . B ATTE LLE -COLUM. BUS LABORATORIES EXHIBIT 8 Page 9 o f 34 HEALTH PHYSICS SERVICES EXTERNAL RADIATION SURVEY REPORT . Building 3 M -l f) Dates om'1 Macht. Area or Room (s) O's ?> M - - NT' h 6 o p tvt Nature of Suspected Activity DN ~ M h (_.- i E CMM PTlb.A ro 5 +4 : p ri N o our L.c O t_o w Mott L Surveyor s N196C.8 Survey Ins trument: type (s) 14 < G.W\\ Give exact location either by writing, attached floor plan, or dbawing on back of this form.
- - Results of Survey z (Units in c/m, rad or(Fnrad/hrQ or n/cm f 3,c ) Reading and Reading and Lo cation ~Tvoe Activity beation Tvoe Activity - . .. ~
, , . . . . . . . . Remarks by Surveyor: 6-4 u d OM4 % PacoL._w r w w w o% g t_ 9t e m uu; M n Prem C hw mh Ng as wmo_$ Remarks by Health Physicist: , . . _. _.. H PS - R S : 6.8_.._, . _ _ _ _ _ __
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EXHIBIT 8 ...!. . - Page 11 o f 34 f.. f.Us Loouaokits . 4' - ' L ATI EL1.E.CU , = . . liEAlJri IHYSICS SEKVICES '
SFCAR StiRVEY PEPORT
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Attachment Ill (2 pages) NRC Confirmatory Survey Record of Transporter and NAC 1-E Cask
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.. .. . .- ' EXHIBIT 8 Page 25 o f 34 , ' . - -- .. . . . . .. t _ CYbrTER C_5EEK NUCt EAR GENE:. ATING STATICN003 % ' RACIATICN FRCTECTION SURVEf RECORD c. ara , 7-25' -fD i LCCATICNi 6P5QT gg)E,(,cA5(, M M DA/ - m:E y s : s s i t. E.:L i= m s.n / I 9 PsS I?DD'I'1 0 , ' , _.,, = I ors:.u.
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' EXHIBIT 9 Page 1 of 1 Dato: July 24, IS U PRELIMitMRY i;0TIFICATIO!10F EVEtiT' OR U;; USUAL OCCURREtiCE--Pl;0-1-80-109 This preliminary notification constitutes EARLY notice of events of POSSIBLE' The information presented is as initial' , safety or public interest sianificance.
received without verification or evaluation and is basically all that is kiiown by IE staff on this date.
Facility: Jersey Central Power and Light Company Oyster Creek (Dri 50-219) , > Forked River, Itew Jersey l l Subject: C0:4TAMl;&TED EMPTY SPEttT FUEL SHIPPIf4G CASK 23, 1980, Gyster Creek received an empty l'odel rio. !;FS-4 spent fuel cask On July The removable radioact:te from BattelJe Menarial Institute of Columbus, Ohio.
contamination on the outside of the cask exceeded both the fiRC 10 One reporting level and the D0{ 49 CFR 173.397(b) limit for trU'nsportation.i;RC reporti smear of 23,000 dp:/100 cm exceeded the 22,000 dp r/100 cm and DDT transportation limit for an exclusive use shipment.
The licensee notified the Headquarters Duty Officer of the reportable contaminatic via emergency notificatica system at 7:10 p.m. on July 23, 1980.
Fedia interest may result due to high visibility of transportation of rodioact'..e The !;RC ar.d material and general interest concerning the Oyster Creek facility.
the licensee do not plan press releases.
The State of !;ew Jersey has been informed of this situation.
Contact: W. W. Kinney, RI 488-1213 R. Keinig, RI 488-1252 , Prepared by EXT SectTo~n Chief ETT" . . E jf. Crocker,_RI 488-1217_ Section Chief EXT
< . PREL!!G!.M'?~t10IIFICATIon ..g't a :. erm ; % l . (F.ev. January 19LO) i f g . ,,, . - '
- . .. , .. , - EXHIBIT 10 Page 1 of 6 A m h- 0-c.18 UA NUCLEIE F.EGU' AY COWL 1LWG84 U 3I ,g gyg 73 CERTIRcATE t CX. 'L1 A54CE i ' For Redk.a.. the 84rnet 7.P 6 ps . 1.Lal Certsf.:ste %rter 1.b) Re '. h::in Flo.
1.(c) P4:as;s L% tWc:rkri No.
1.fdl Pys M. 1.f e) Tctag *so. r, pas 6598
USA /6603/B( )F
6 2. F R E Au b tE 21:1 TMs es ttGer e h issue to actis +r Sr-hrs 1712.3'.a.173.54.173f 5.
i:f 173 2:'* of tt.* C=;wienemt of Trtso staan Hes.som C Wic-!s's F ;.crjons (d3 CF F.170-1C3 e.=d la CFR 1C 3) at S..kns its--M-1^.-a are 1cf -19-17". of tN D.catment et Trucor1r:5n Carge-om Crryxs F.c.sdath:ss M3 OFR 14 "-10). as arr,was.
' * tb) TM WW.ng e 3:f ccma.nts de-:.= bef h iterii s t e rv. rnn s st-s :hty semeures : fxth h 5 :pm C of Titf e 10. Coce of . Fede r4 As; satio.s. Ps t 71. *'Pa.m of Re$une V.-wid for Trm::;x>rt e-w! Tserc.x:rriri n of FvEcc::tive Mrtertst U cf ar Cc-ina Corti's soris "
, 2.fc1 nns certin= e o>es ros r:4 t t'se cors); or frarn ccT:$:r.cz with ar,y rs:;ui.wt of t?.e resv'-es-a of 1.% U 5. o-E.*-twens of Tren.oort:tions or other arc 41cds rrW:iry a r-di. bef tr:f 2 the c:vserr w,1 e' or,y cou, y thru;S or in v wS*.ch t% ;ncupe e. tre.-e.
3. Th:s c:rtl**cr.e is I tal en the t ?:' c,f a s-fry e.97is r.M cf ?
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33) T;f.c or.3 &t:t.-.u-bn of rt; :* car s:-S=**c.n- .Lclear Fuel Services, Inc.
NFS Epplication dated October 6,1972, P.O. Box 124 c.s srppic :nted.
+ Ucst Valley, NY 14171 2.te) c.-o.s ra.71-6698 4 CCf.?tTIONS This tr-tife=:e is cori:Wca.21 Upon t**g f fl!i g of the rt2uirr~c-s of E t.owt O of to CFR 71, sa opr.cdde, end the cc9deicr s : eGus . - in h: b S W tes.
F' 1,. oe.:.nc ;c.a of emina -.s us, ore::e ce r.-<-:3. % no,e.r. ra:-7, ca. o-,,., Ce,.c2tw. ce e_:-r, c.s.
(a) Packi:ging (1) Model No.: NFS-4 (2) Description A steel, lead and water shielded shipping cask.
The cask is a right circular .' cylint.r with upper and lower steel er cased balsa impact limiters. The overall dimensions are 214 inches in length and 50 inches in diameter. The l gross weight of the cask is approxina.tely 50,000 pourids.
The inner cavity l is 1.78 it % long and 13.5 inches in diameter.
The thickness of the inner shell is 5/io inch and 1-1/4 inches for the outer shall.
it.e two stainless I steel shells are, welded to a 2-inch thick stainless steel shield disc at the bottom.
The annulus between the inner and outer shells is filled with i lead (cax. lead thicknes: 6-5/8 inches, minimum 5 inches).
l i The lid is s,tainlf:ss steel frustum of cone 7.5 inches thick.
The lid is secured to the cavity flange by six ASTM-A320, Grade L43,1-1/4 inch dia.T.eter bolts.
The seal is provided by two polytetrafluoroethylene 0-rings.
Four , l neutron shield tanks, cach with surge tank and rupture disc, provide a j. 4-i/2 inch thick (borated) water-ethylene glycol mixture.around the outer ' shell. Four trunnions, two located on either side of the upper or lower impact limiter, are provided.
Other cask features include two drain valves
located in the bottom shield disc, vent valve, head closure gasket leak check valve, ruptu're disc-pressure relief valve system located in the " cavity flange, fuel canisters for WP, and BWR shipments, and spacers to accor:r::odate shorter fuel asser.blies.
For transport the cask may be encle'ed in an expended metal cage.
.
- . , . . .- - EXHIBIT 10
Page 2 of 6 , Page 2 - Certificate t;o. 6698 - Revision No. 9 - Do:}.et rio. 71-6598 h
5.
(a) Packaging (centinued) . ~ (3) Drawings The. NFS-4 shipping cask is constructed in accordance with Nuclear Fuel 'f ' Services, Inc., Drr. wing No. E 10',30, P.ev.19 (Shaets 1 through 4). The ,y . 'uel assemblies are p3sitioned within the fuel canisters sh:wn in Figure 2.1.3 of the application dated G:tober 6,1972.
Spacers may be used to . acco=_v.! ate shortar fuel assechlies within the fuel canisters.
, (b) Contents - , (1) Type and form of raterial The minir.ua cooling tire of each fuel issc-bly and rod shall te 120 days.
- and (i) Irradiated ph2 or EUR uranium oxide fuel assemblies with the following . maximum active dimensions and taximum compositions prior to irradiation: Fuel Asstrably Data PUR Bh'R Envelope, inches 8.60x3.60x150 5. 44x5.44x144 , Enrichment, w/o U-235 3.6 3.0 , Weight of Uranium, kg 430 197 H/U atcaic ratio - 5.51 ' (ii) Fuel asse;nbly enriched in the U-235 isotope to not c. ore than 2.h w/o, with octive fuel dim 2nsions r.ot to exceed 4.2" x 4.2" x 110" long.
(iii) Byproduct and special nucisar caterial in the fem of irradiated . uranium oxide fuel rods.
- (iv) Solid nonfissile irradiated hardare and neutron source components.
(v) Fuel asse.mbly enriched in the U 7'i5 isotope to not Fore than 4.1 w/o, with active fuel dirrransions not to' exceed 7.8" x 121" long.
. (vi) Byproduct and special nuclear material in the form of irradiated uranium and plutonium oxide fuel rods.
Prior to irradiation, the m:.ximum enrichment in U-235 plus plutonium not to exceed 4.0 w/o.
(vii) Irradiated Ph1 uranium oxide fuel asstablies including additional irradiated fuel rods inserted and secured in the guide thimbles.
The - fuel asserblies shall confcm to the maximum active ~ dimensions as _ described in Item 5(b)(1)(1) and partially disassenbled fuel asse::blies ,.. . shall be equipped with an asse-bly carrier as shonn in Battelle " , Drawing Ho. 00-001-676, or equivalent.
.
e f .
l EXHIBIT 10
. , . . Page 3 of '6 - Fage 3 - Certificate No. 6598 - Revision ilo. 9 - Docket H'o. 71-6598 e . s ' 5.
(b) Contents (continued) . (1) Type end fcrm of taterial (continued) Prior to irradiation, (viii) Irradiated S'dR' uranium oxide fuel assemblies.the maxie.un enrichme with active fuel dimensions not to exceed 5.63" x 5.63" x S3.8" long.
(2) P.axim.:a quantity of r.aterial per package.
. Not to e. ceed a decay heat genaration of 2.5 kw and . (i) Item 5(b)(i) cbove: - One (1) PG fuel assc.bly, or Tito (2) BG feel essc.blies; or (ii) Iten 5(b)(1)(ii) above: - Four (4) fuel esst blies contained within the fuel basket shown in NFS 0 g. No. l A-T-1107, Rev. 0; or (iii) Iten 5(b)(1)(iii) above: . ,. h.-- ' tisxicam Fissile .. ass l_imit , " thximum Enrichment (w/o U-235) (ho of U-235) '
2.0
1.6 -
1.5; or !! (2) P.eximum quantity of traterial per package (continued) i.
(iv) Iter.5(b)(1)(iv) above: ~ l . r . ' As r.eeded, appropriate component spacers shall be used in the cask cavity to limit novement of contents during shirmsnt; or (v) IttA'5(b)(1)(v) above: ' . One (1) fuel assembly; or (vi) Iten 5(b)(1)(vi) above: Fuel rods with.a the fuel canisters described in 5(a)(3).
The iraximum mass of U-235 plus plutonium shall not exceed 4.0 kg.
A suitable - l fixture may be 21 sed to secure the fuel rods within the ccnister; or . i -- ' q
- t . . ! " ~ - , -
EXHIBIT 10
- , Page 4 of 6 , Page 4 - Certificate No. 6593 - Kevision No. 9 - Docket No. 71-6698 h9 . 5.
'(b) Contents (continued) (2) Maxirm quantity of r,aterial per package ' continued) ( (vii) Item 5(b)(1)(vii) above: The iaximum compositions of one FiiR fuel assembly including additio.al rods shall conform to Item 5(b)(1); or . ' (viii) Item 5(b)(1)(viii) above: _ Two (2) BUR fuel assemblies.
prior to irradiation, the maximum ' uranium content per assably shall not exceed 122 kg.
(c) Fissile Class III P.aximum nufaer of packages par shipc.cnt One (1) ~ 6.
The cask shall be shipped dry (no water coolant in cask ccvity).
j 7.
The water-ethylene glycol mixture in the neutron shield tanks ray contain up to 1.0 s. sight percent baron.
This mixture shall mt freeze or precipitate in a temperature range from -40*F to 330 F.
( 8.
The cask contents shall be so limited under norral conditions of transport that 27 times the neutron dose rate plus 1.4 times the garns dose rate will not exceed 1,000 millirems per hour at three (3) feet from the external surface of the package.
9.
The vent'and drain valves shall be 1/2" FG466T5W Hiser ball valves (Worchester Valve company,Inc.). The ball of the valve rray have a bleed hole to equalize the pressure between the cask cavity and the ball passage in a closed position.
Alterna tiv ely, the vent and drain lines may be sealc4 with pipe plugs.
' j 10.
In addition to the requirements of Subpart 0 of 10 CFR Part 71, each package prior to first use shall meet the acceptance tests and criteria specified on pages A-21 thru A-34 of the Nuclear Fuel Services, Inc. application dated October 6,1972, as atended, f: arch 1,1973 and Nuclear 4.;urance Corporation letter dated November 1,1974.,The results of these tests shall be docu.ented and retained for the life of the cask.
11. At periodic intervals not to exceed (3) years, the thental perforrance of the cask shall be analyzed to verify that the cask operation has not degraded below that which is licensed.
Following the initial acceptance tests, the heat source may be that provided by the decay heat from the contents of the package provided that the heat source is equal to at least 25% of the design heat load for the package.
Each cask that fails to meet the therr.al acceptance criteria given on pages A-21(a) and A-21(b) using the TAP craputer program, or equivalent, shall be withdrawn from service until corrective action cr.n be completed.
, .12.
The rupture discs for the neutron shield tanks shall b': type "B" or "DV" (BS&B Safety - Systems, Inc.) or equivalent.
. .
EXHIBIT 10
. . Page 5 of 6
Fage 5 - Certificate No. 6598 - Re ision No. 9 - Docket No. 71-5598 h - 13.
In lieu of the requiremsnts of 10 CFR 571.54(h), the licensee shall perform periodic maintenance and testing of 0-rings, drain and vent ball valves, relief valves, and rupture discs of the cask as indicated in the table given below.
Curing inactive periods, the maintenance and testing frequency may be disregarded provided that the package,is brought into full compliance prior to the next use of the package.
Cask Ccmoonent Period Test / Action . Ball Valve ? Each shipm.ent Hydro test to 80 psig*
- Ball Valve Annually Replace seats and seals
0-rings Each shipr.ent Test to 80 psi * S - 0-rings Quarterly Test to 167 psig* 0-rings An'nually Test to 1005 psig* Inner Containaent Vessel Quarterly Test to 250 psig*
Cavity Relief Valve Annually Test at set point , Cavity Rupture Disc An'nually Replace Neutron Shield Tank _ Rupture Disc Annually Replace , b Impact Limiters Annually , Test for leakage =There scali be no visual (pressure gauge) indications of pressure drop for the component under test during a 10-minute test period.
Otherwise, corrective action shall be taken and the test repeated until such time as the cor.,r.onent meets the specified test.
(Test to pressures equal to or greater than those indicated.)
. I 14. At least every six (6) months, co.nencing six months following the first shitment of irradiated fuel under Revision No. 9 to Certificatt-of Ccmpliance No. 'S8, the licensee shall perform physical measurements of the .sk inner shell.
Any cask ecse inner container dimensions are measured to deviate by more than +0.015 inch l at comparable points from the dimensions docunanted in Appendix C to NEC letter
dated June 8,1979 shall be removed from service.
. 15.
The package authorized by this certificate is hereby approved for use under the general license provisions of 10 CTR 571.12(b).
16.
Expiration date: December 31, 1980.
' i
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' I I . . ! [ -
V . . .. . ' ' ' EXHIBIT 10
- . Page 6 of 6 Page 6 - Certificate No. 6593 - Revision No. 9 - Docket tio. 71-6598 - - . REFERENCES ' riuclear Fuel Services, Inc. application dated cctober 6,1972.
Suppler.ents dated: November 9,1972; January 10 and 22 February 1 and 28, Farch 1,14, tnd 21, F.ty 4, June 4, and July 26,1973; July 17,1974; t'ay 4,1976; and November 9, 1977.
!!uclear Assurance Corporation supplements dated: ricver.bar 1,1974; August 13 and Decenber 24 197 ; September 13, 1976; October 20,1977; l'ay 25, July 18, and September 25,1978; and
June 8, July 25, and Octat>er 31, 1979.
. , FOR THE U.S. 'iUCLEAR REGULATORY CC+ MISSION . f fa.h [A C-v Charles E. FbcDor,ald, Chief . Transportation Certification Branch Division of Fuel Cycle and Interial Safety cate: - DEC 12 iM . $
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' $ffk b.RFoV . EXHIEIT 12 Page 1 of 1 CHRONOL CC, V . NO7 Cnic nosr pensonsto t-s-ao en este pErfonneo wirH 1971 ANS PAopoSED S-22-80 US/MG srnoonRD nas /979 sinwpMRD (topa ucreo/> )- wird . l Aws: 2.of kW VncrRrd/NTIES AMS : 3.5/ kW Q 4 MAC cate 5-27-80 % AMS.- 2.97 kW max ROO TEMP EQVM ro S70"f (S2S*F G 2.SkW) A Barrzzie O9ic - nos: 3.07kw pva-to sliu,In o.
G ll CALC F/AST PERfQ9MED TVSr QA cnte PERFogmED PAto2 70 SM/PoATE (S/7{Bo) 5-30-80 US/MG AMS /979 wirH l97/ MNS PROPOSED srAMDMAD srMHDMnD ( simpuriCO METHOD) - Wird uxcerrmanes Inos:.78kw} a.37 kw y 'nas:
- Bntreue Cntc AMS: ~/[NW l
Soot / cAtt nasr prproaxeo susr GR entc prnfonMED PRiun to sniponte (r//4/Pz>) 5 30-90 yyns dos /971 ! wirp nas /979 sindantD srAwanno l (Simpitfico MErHoD)- NC (S/MPLIF/CD M/hD) ~ LUtrW udcEniniancs uccEnrninrics nos : . 24 kW k lryys: /. / kW Q t x l l
_ - ' ' .. - , EXHIBIT 13 Page 1 of 7 (~ OYSTER CREEK NUCLEAR GENERATING STATION 925T) h
l RADIATION PROTECTION SURVEY RECORD DATE W-5~-ko . Sff y pggL g,73gi 7,acjf LOCAT10N: ME i l'Y30 REASON FOR SPECIAL SURVEY: M4 /2 ROUTINE C SPECIAL O ! . rrEus DCSE RATES OtSTANCE CONT NA ON SMEAR "fe" l I '_l.,J I C I.'".IZ," I tI ...m , ti ' lNn i 4 % l N il l & 2. I I \\< / K \\tsav it/ W M , i
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N #/ W 2W7 SURVEYOR X< 3d REVIEWED BY d' , ' Form 1542 l - _ _ _ _ _ _ _ _. _ _ .... _ _ _.. _ _ . . . . .. -
_ _. ' EXHfBIT 13 - - ,.: Page 2 of 7 l(' OYSTER CREEK NUCLEAR GENERATING STATION 4747G Y
' RADIATION PROTECTION SURVEY RECORD DATE8-5-ko
LOCATION:$f d f h g ' TIM ! REASON FOR SPECIAL SURVEY: II ' ROUT 1NE C ' ,' SPECIAL f/ \\ l cSEf/EEcu
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- EXHIBIT 13
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_ _ _ _ r 3) EXHZBZT 13 ' - Page 5 of 7 - .. , (- N o.
OYSTER CREEK NUCLEAR GENERATING STATION 4),129-/s RADIATION PROTECTION SURVEY RECORD DATE ' . YSkD LOCATION: //f2 [/gr /'g~ e f"'de '7 ~ TIME ' / N./fL-(%S2 NO REASON FOR SPECtAL SURVEY: ROUTINE E-SPECIAL G rrEus DCSE RATES DISTANCE C NT NA CN SMEAR "# I l'l,:I IMI I t1; .m 1i i l i i i 1 M2rv l e/0TI
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I EXHXBIT 13 P_a g e 6 o f 7 ,, o i l (. OYSTER CREEK NUCLEAR GENERATING STATION 7-79 ' RADIATION PROTECTION SURVEY RECORD DATE LOCATION: //f [g/f @,p fgpg TIME # fc=.s v 6.rz-REASON i:OR SPECIAL SURVEY: ROUTINE K ' ' SPECIAL O . i.
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