IR 05000354/1986045
| ML20214F821 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/17/1986 |
| From: | Kottan J, Nimitz R, Paolino R, Shanbaky M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20214F713 | List: |
| References | |
| RTR-NUREG-0737, TASK-2.B.3, TASK-2.F.1, TASK-3.D.3.3, TASK-TM 50-354-86-45, NUDOCS 8611250448 | |
| Download: ML20214F821 (33) | |
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- U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. 50-354/86-45
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Docket N License N NPF-50 Category B Licensee: Public Service Electric and Gas Company 80 Park Plaza, 17C Newark, New Jersey 07101 Facility Name: Hope Creek Generating Station Inspection At: Hancocks Bridge, New Jersey Inspection Conducted: September 22-26, 1986 Inspectors: bb 11 k 19M R. L. Nimitz, Senior Ridiation Specialist date (Team Leader)
NJ & ll17 Gb J. J. Kottail, Littio ~ t y Specialist I dhte
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f lR. J. Paolino, L'ead Reactor Engineer I date O A. P. Hull, Brookhaven National Laboratcry Approved by: W .% A, /dv #//7/4 M. M. Shanb'aky, Chief,4acilities date Radiation Protection Section Inspection Summary: Inspection on September 22-26, 1986(Report N /86-45)
Areas Inspected: Special, announced safety inspection of the licensee's implementation and status of t5e following task action items identified in NUREG-0737: II.B.3, Post Accident Sampling Capability; II.F.1-1, Noble Gas Effluent Monitors; II.F.1-2, Sampling and Analyses of Plant Effluents; II.F.1-3, Containment High-Range Radiation Monitor; III.D.3.3, Improved Inplant Iodine Monitoring. Three region based inspectors and one contractor from Brookhaven National Laboratory participated in the inspectio Results: No violations were identifie Several areas requiring improvemert were identifie PDR ADOCK 0500 G
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INDEX 1.0 Individuals Contacted 2.0 Purpose 3.0 TMI Action Plan Generic Criteria and Commitments 4.0 Post-Accident Sampling System - Item II. .1 NUREG-0737 Item II.B.3 Position 4.2 Documents Reviewed 4.3 System Description and Capability 4.4 PASS Performance Testing 4.5 Reactor Coolant Sampling 4.6 Containment Atmosphere Sampling 4.7 Analytical Capability 4. Chloride 4. Boron 4. pH 4. Radioactivity 4. Hydrogen and Dissolved Gas ,
4.8 Additional Findings 5.0 Post Accident Sampling - Item II.F. .1 NUREG-0737 Item II.F.1.1 Position 5.2 Documents Reviewed 5.3 System Description and Capability 5.4 Acceptability 6.0 Post Accident Sampling - II.F.1-2 6.1 NUREG-0737 Item II.F.1.2 Position 6.2 Documents Reviewed 6.3 System Description and Capabilities 6.4 Acceptabilr-j 7.0 Containment High Range Monitor - Item IIF.1-3 7.1 Position 7.2 Document Reviewed 7.3 System Description 7.4 Findings 8.0 Improved In-Plant Iodine Instrumentation Under Accident Condition -
Item III D. .1 Position 8.2 Review Criteria and Documents Reviewed 8.3 Description of Methodology and Capabilities 8.4 Findings
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9.0 Post Accident Sampling System - Quality Assurance and Design Review
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9.1 General 9.2 Dacuments Reviewed 9.3 Installation 9.4 Findings 10.0 Licensee Commitments 11.0 Exit Meeting b
12.0 Attachments:
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- Attachment 1- Individuals Contacted
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- Attachment 2- Documentation Associated with NUREG-0737 Item II. * Attachment 3- Comparison of Chemical and Radiochemical Test Results
- Attachment 4- Documentation Associated with NUREG-0737 Item II F.1-1 and II F.1-2 Attachment 5- Documentation Associated with NUREG-0731 Item II F.1-3 Attachment 6- Documentation Associated with NUREG-0737 Item III D. Attachment 7- Documentation Associated with Design, Construction and Installation of the Post-Accident Sampling System
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DETAILS 1.0 Individuals Contacted The individuals contacted during this inspection are listed in Attachment 1 to this inspection repor .0 Purpose The purpose of this inspection was to verify and validate the adequacy of the licensee's implementation of the following task actions identified in NUREG-0737, Clarification of TMI Action Plan Requirements:
Task N Title II. Post-Accident Sampling Capability
. II.F.1-1 Noble Gas Effluent Monitors II.F.1-2 Sampling and Analysis of Plant Effluents II.F.1-3 Containment High-Range Radiation Monitor III.D. Improved Inplant Iodine Instrumentation under Accident Conditions As part of the inspection, a review was performed to verify and validate the adequacy of the licensee's design and quality assurance program for the design and installation of the Post-Accident Sampling System (PASS).
3.0 TMI Action Plan Generic Criteria and Commitments The licensee's implementation of the tath actions specified in Section 2.0 were reviewed against criteria and commitments contained in the following documents:
NUREG-0737, Clarification of TMI Action Plan Requirements
Generic Letter 82-05, letter from Darrell G. Eisenhut, Director, Division of Licensing (DOL), NRC,-to all Licensees of Operating Power Reactors, dated March 14, 1982
NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations, dated July 1979
Letter from Darrell G. Eisenhut, Acting Director, Division of Operating Reactors, NRC, to all Operating Power Plants, dated October 30, 1979
Letter from Darrell G. Eisenhut, Director, Division of Licensing, NRR, to Regional Administrators, " Proposed Guidelines for Calibration and Surveillance Requirements for Equipment Provided to Meet Item I..F.1., Attachments 1, 2, and 3, NUREG-0737," dated August 16, 982
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- Regulatory Guide 1.3, " Assumptions Used for Evaluating Radiological Consequences of a loss of Coolant Accident for Boiling Water Reactors"
Regulatory Guide 1.97, Revision 2, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following en Accident"
Regulatory Guide 8.8, Revision 3, "Information Relevant to Ensuring that Occupational Radiation Exposure at Nuclear Power Station will be As Low As Reasonably Achievable"
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Final Safety Analysis Report, " Hope Creek Nuclear Generating Station" NUREG-1048, " Safety Evaluation Report related to the operation of Hope Creek Generating Station" 4.0 Post-Accident Sampling System, Item II. .1 Position NUREG-0737, Item II.B.3, Specifies that licensees shall have the cap-ability to promptly collect, handle, and analyze post-accident sam-ples which are representative of conditions existing in the reactor coolant and containment atmospher Specific criteria are denoted in commitments to the NRC relative to the specifications contained in NUREG-073 .2 Documents Reviewed The implementation, adequacy and status of the licensee's post-acci-dent sampling, monitoring, and analysis systems were reviewed against the criteria identified in Section 3.0 and in regard to licensee letters, memoranda, drawings and station procedures as listed in Attachment 2 of this Inspection Repor The licensee's performance relative to these criteria was determined from interviews with the principal personnel associated with post-accident sampling, reviews of associated procedures and documenta-tion, and the conduct of a performance test to verify hardware, pro-cedures and personnel capabilitie .3 System Description and Capability The licensee has installed a Post Accident Sampling System which is a standard General Electric design. It has the ability to obta M unpressurized undiluted and diluted samples of reactor coolant from the jet pump and the RHR System. Also, samples can be obtained from the drywell, suppression pool and reactor building atmosphere Redundant containment hydrogen analyzers provide a hydrogen analysis back-up capability.
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Analysis for chloride, boron, pH and hydrogen are conducted in the laboratory using an ion chromatograph, an ion speciffc electrode, a micro electrode, and i gas chromatograph respectively. Radioactivity analyses are performed in the licensee's counting room using a com-puter based gamma spectrometer. Chlorida analysis can also be per-formed by an offsite laborator .4 PASS Performance Testing Grab samples of reactor coolant and the reactor building (secondary containment) atmosphere were collected during an operation test of the PASS system on September 24 and 25,1986. During this test, lic-ensee personnel demonstrated the integrated ability to collect and analyze samples within the constraints of NUREG-0737, II. .5 Peactor Coolant Sampling The reactor coolant sampling system is designed to obtain samples of liquids and dissolved gases during all modes of operation. During this operational test samples were collected from the RHR system be-cause the reactor was shutdown. Although both liquid and dissolved gas samples could be obtained from the prescribed sampling points, the following improvement items were discussed with the license The licensee stated that these matters will be reviewed and clarifi-cation / improvements will be considered, as appropriate: (50-354/
86-45-01)
a During the first attempt at taking an undiluted reactor coolant sample, the sample needles were bent and no sample was obtaine When a second attempt was made to obtain an undiluteo reactor coolant sample, the sample vial did not contact the position indicating switc The licensee indicated he would review and evaluate the sample vial positioning and alignment and make any necessary changes needed to ensure sample collection
The liquid sample vial septum retaining ring was damaged during sampling (split) to the extent that it no longer retained the septum. This can result in sample loss during shipment to an offsite laborator The licensee indicated he woald evaluate the septum sealing mechanisms and make appropriate septum sealing / vial changes if
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. 5 4.6 Containment Air Sampling Atmosphere samples can be obtained from the drywell, reactor building and suppression pool. During this operational test, samples were collected from the reactor buildin The following items needing clarification or improvement were identified: Th.e licensee indicated these items would be reviewed for clarification or improvement (50-354/86-45-02)
- The licensee requires collection of airborne iodine on silver zeolite cartridge However, charcoal cartridges were used for the tes Suitable silver zeolite cartridges were not available at the Hope Creek site but were available at the neighboring Salem site. Also the cartridges were not purge The licensee indicated he would review the need for purge cap-ability for the containment air sample and use silver zeolite cartridges in the sampling syste *
The licensee was unsure of the method used to calculate the air-borne particulate and iodine activity: whether to use the air sampler rotameter reading or the flow maintained by the limiting flow orifice. The PASS sampling procedure does not specify how to perform the calculatio .
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The l_icensee indicated he would establish clear guidance fo determination of containment airborne particulate and iodine
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Tha PASS sampling procedure requires that an airborne particu-late and iodine sample be taken for a preset time rather than allowing the technician taking the sample to vary the time based on the amount of radioactivity being collected on the sampling media. This may lead to unnecessary handling and analyzing samples with high activit The licensee will consider revising procedures to allow minimi-zation of maximum dose rate on sample cartridges by limiting
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sample collection time but still allowing representation samp-lin .7 Analytical Capability The licensee's commitments relative to range, uncertainty and analy-tical capability are contained in his Final Safety Analysis Report (FSAR). The Safety Evaluation Report (NUREG-1048) specifies that the accuracy, range, and sensitivity of the PASS Instruments and analytical procedures are consistentwith R.G 1.97, Rev.3 and NUREG-073 ._. . - - _
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4.7.1 Chloride The licensee's primary method for chloride aralysis is the use of an offsite laboratory, B&W's Lynchburg, Virginia laboratory, to provide results on an undiluted sample within four days. The licensee can also perform chloride analysis in-house using an ion chromatograp The licensee indicated the following matters would be reviewed for clarification or improvement (50-354/86-45-03). The licensee was found not to be a registered user of the shipping cask which was designed for shipping out PASS un-i diluted liquid samples. Also no procedures were establish-ed for sampie ioaoirg, and clesure for the cask. The lic-l ensee could, however, ship the samples out in other Type B casks but was primarily relying on the PASS cask for this purpos The licensee indicated he would identify what means will be used to ship a PASS sample to B& The licensee has not established maximun dose rate criteria for use in determining when a PASS liquid sample can be analyzed in-house without exceeding GDC-19 criteria. Tae licensee should establish limiting dose rate criteria for in-house analyses of PASS samples for chloride Chloride standards were submitted to the licensee for analysis in-house and by the offsite laboratory. The re-suits are listed in Attachment 3. The licensee's analysis results were acceptabl .7.2 Boron Boron analysis is performed using a fluoroborate specific ton electrode in the licensee's laboratory on a diluted reactor water sample. Boron standards were submitted to the !icensee for analysis. The results are listed in Attachment 3. The lic-ensee's analysis results were acceptabl .7.3 pH Analysis for pH is performed using a microelectrode in the lic-ensee's laboratory on an undiluted 0.3 mi sample. Results on an actual sample are contained in Attachment These samples can be analysed in the laboratory. The licensee's analysis results were acceptabl l
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4. Radioactivity Analysis Gamma isotopic analysis of PASS liquid and gaseous samples is performed using the licensee's normal counting room gamma spec-troscopy syste The use of dilution permits the licensee to analyze the full range of anticipated concentrations in liquid sample Airborne iodine and particulate samples up to approximately one millicurie of total activity can be analyzed. The sample flow can be adjusted to limit the amount of radioactivity on the filter and cartridg Larger amounts of iodine on a cartridge can be assessed using a dose rate instrument with appropriate conversion factor Results of an actual reactor water sample are contained in Attachment The licensea's analysis results were acceptabl The licensee indicated the following items would be reviewed clarification or impovement:
The licensee has established and implemented procedure CH-TI-ZZ-Oll, " Estimation of Reactor Core Damage Under Accident Condition". However, comparison of the particular
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radionuclide concentrations required to be known to ~-
implement this procedure indicates that the radionuclides
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are not all determined via procedure CH-RC-ZZ-004, " PASS Post-Accident Sample Analysis". The licensee will review his core damage assessment methodology to ensure all appropriate radionuclides are determined inorder to im-plement procedure CH-TI-ZZ-01 (50-354/86-45-04)
4. Hydrogen and Dissolved Gas Dissolved gas is determined by the GE PASS expansion metho This method is used to quantify hydrogen. This method was demon-strated by the licensee. Hydrogen and oxygen analyses can also be performed on samples by gas chromatograph The analysis of hydrogen in the containment atmosphere is pro-vided by an in-line hydrogen analyzer as required by II. F.1- The licensee indicated that the following matters would be re-viewed for clarification or improvement:
The procedure for perforning dissolved gas sampling at the PASS did not provide guidance for re-initiating sample flow following the collection of a liquid sample. Within re- .
initiation of sample flow no analysis can be provid? However, the technician performing dissolved gas sampling at the PASS was able to re-initiate flow and acquire the required sampl .
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The licensee indicated he would revise the dissolved gas portion of the PASS procedure and retrain and qualify technicians on the procedur (50-354/86-45-05)
Note: The licensee provided some commitments in this area to ensure his capability to perform dissolved gas sampling prior to reactor start-u (See Section 10 of tnis report)
4.8 Additional Findings The licensee indicated the following items would be reviewed for clarification or improvement (50-354/86-45-06): -
The licansee technicians, which were assigned to perform the PASS sampling, were not qualified on the use of self contained breathing apparatus. Technicians may need to wear SCBA The licensee indicated he would train and qualify selected chem-1 try personnel on use of.SCBA *
Dose limits have not been established for removing from the transpcrt shield and handling an undiluted vial of reactor water ir the chemistry laborator tie licensee indicated he would establish maximum allowable dose rates for handling of PASS undiluted liquid samples to ensure meeting GDC-19 criteri *
The tongs used for removing the containment gas sample vial from the sample vial holder do not work. The vial was removed by turning the holder upside down and hitting it against the floor until the vial fell out. This can result in vial breakage and loss of the sampl The licensee indicated he would establish an acceptable methodology for removal of the gas sampl .0 Noble Gas Effluent Monitor, Item II.F. .1 Position NUREG-0737, Item II.F.1-1 requires the installation of noble gas mon-itors with an extended range designed to function during normal and accident conditions. The criteria, including the design basis range of monitors for individual release pathways, porer supply, calibra-tion and other design considerations are set forth in Table II.F.1-1 of NUREG-073 ,
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c 5.2 Documents Reviewed The implementation, adequacy, and status of the licensee's monitoring systems were reviewed against the criteria identified in Section and in regard to documents listed in Appendix The licensee's performance relative to these criteria was determined by interviewing the principal persons associated with the design, testing, installation and surveillance of the high range gas monitor-ing systems, reviewing associated procedures and documentation, exam-ining personnel qualifications and direct observation of the system .3 Description and Capability The station has three gaseous effluent release paths, the North Plant '
Vent (NPV), the South Plant Vent (SPV), and the Filtration, Recir-culation and Ventilation System Vent (FRVSV). the NPV and the SPV discharge points are slightly above building rooftop at an elevation of approximately 115' while the FRVSV has an independent effluent line which discharges at the top of the containment building at an elevation of approximately 200' .
The NPV and SPV are monitored for routine gaseous effluents by a General Atomic (GA) RD-52 beta scintillation detector which is a com-ponent of the installed (GA) particulate, lodine and Noble gas (PIG)
monitors. All three paths are equipped with GA Wide Range Gas Moni-tors (WRGM). Those for the NPV and SPV are provided with GA Rd-72 mid- and high-range Cd-Te detectors. The WRGM for the FRVSV is also provided with an RD-52 low-range beta scintillation detecto GA RM-23 control and readout nodules for all three WRGMs are located in the Control Room. These modules can display all monitor para-meters, including channel activity, flow rates, alarm status, check source actuation and purge status. Chart recorders of concentration and release rates are also provide The GA-RD-72 Cd-Te detectors, are sensitive to both are beta and gamma radictions. The high-range capability is established by an approximately 100 fold reduction in the volume of gas viewed by the high-range detector (approximately 30 cm3 vs. approximately 0.2 cm3 ).
At the present time, raw data of the concentration of radiogases as indicated by the RM-23, is utilized in the licer.see's dose assessment
. procedures. However, the licensee's centractor, has developed a scheme for the correction of the WRGM's readings to total gas con-centrations as a function of the time post shut-down, for a changing post accident mixture of radiogases. A computer program, CRCONV, has been written to perform the time-dependent response calculation The licensee has accepted GA's original type calibration of the RD-72 detector. However, GA's calibration report (e-225-961, Rev. 3) does
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not indicate what source concentrations were used at the time of the original calibratio Determinations of the linearity and energy response of the original detectors and of the efficiency of subseq-uent production RD-72 detectors have been performed by GA using transfer solid sources which are NBS traceabl The high-range channel of the WRGM has a very small volume which is established by a compression fit of the detector against a washer in a small recessed cavity of the detector shiel It was uncertain to the inspector that the fixed sample / detector geometry, which is util-ized in the calibration procedure, accurately reproduces that of the detector when installed in the high-range position in the shiel The licensee has not established a list of spare detectors, sensors, pumps and spare boards for the RD-11 and RD-80 monito .4 Acceptability Providing that the rationale for the conversion of monitor readings can be clearly established and its accuracy assured, the system as reviewed would meet the requirements of NUREG-0737 Attachment I F.1- Within the scope of this review, the following items needing licensee clarification or improvement were identified. (50-354/86-45-07) The licensee indicated these items would be clarified or improved:
The logic of the CRCONV program to correct detector response to total concentration appears to be conceptually correct. How-ever, its use requires an additional correction in the dose assessment procedur The licensee will consider the RM-23 readout either directly as Xe-133 equivalent or interpretable as such through the use of appropriate correction factor *
The licensee indicated the concentrations of 233Xe and Kr at which the vendor performed the original type test of the RD-72 detectors would be verifie *
The licensee indicated the linearity of the RD-72 detectors as they approach maximum concentration of 10' uCi/cm3 would be verifie *
The licensee indicated the reproducibility of the volume viewed by the high-range detectors as installed in the shield would be verifie *
The licensee will consider the regular verification of data set points in the proposed revision of RP-AP-SP001(Q)
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The licensee will consider the need for a list of reou b ed soare parts for the GA RM .0 Sampling and Analyses of plant Effluents, Item II.F.1-2 6.1 position NUREG-0737, Item II.F.1-2. requires the provision of a capability for the collection, transport, and measurement of representative samples of radioactive iodines and particulates which may accompany gaseous effluents following an acciden It must be performable within specified dose limits to the individuals involve The criteria including the design basis shielding envelope, sampling meoia, sampling cons 1cerations, and analysis considerations are set forth in Table II.F.1- .2 Documents Reviewed The implementa'. ion, adequacy and status of the licensee's sampling and analysis system and procedu-es were reviewed against the criteria identified in Section 3.0 of this report and in regard to licensee correspondence, memoranda, drawings and station procedures as listed in Attachment ~
4The licensee's performance relative to these criteria was determined ~
by interviewing the principal persons associated with the design, testing, installation, and surveillance of the systems for sampling and analysis of high activity radiciodine and particulate effluents, by reviewing associated procedures and documentation, by reviewing personnel qualification, and by direct observation of the system Description and Capabilities As discussed in section 5.3, the licensee has three possible airborne release pathways; the NVP, the SPV and the FRVSV. To fulfill NUREG-0737 requirements regarding particulate and iodine sampling at low conc.ntrations the NPV and SPV are provided with a moving tape, part-iculate filter and fixed iodine filter which are viewed by NaI de-tectors in the PIG unit. The requirement for sampling of post accid-ent concentrations is met by the installed mid- and high-range samp-ling filters in the WRGM' An isokinetic sampling flow is maintained to the NPV, SPV and FRVSV monitors. All sample lines are heat traced and insulated. Vent and sample mass flow measurement techniques are employe The FRVSV WRGM contains a feature by means of which filters can be purged prior to their removal and replacemen _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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During a simulated sample change-out at the FRVSV sample skid, it was observed that about 10 minutes is required to open the lead shield and to extract the particulate / iodine cannister. This is done for icw dose rate samples. The complete shield can be removed if the collected sample exhibits high dose rates. The licensee estimated that this would require about forty-five minutes. Although a local radiation monitor is provided for the FRVSV skid area, its readout is not visible to persons during sample changeout.
Under worst case, a deposition of elemental iodine of 76.9% would occur in the FRVSV skid at the minimum isokinetic sample flow rate of 0.025 cfm. Corrections are made for deposition fractions in appli-cable procedure .4 Acceptability The system as reviewed meets the guidance issued by the NRC in NUREG-0737, Attachment II.C.1- Within the scope of this review, the following items were identified i which the licensee indicated would be clarified or improved:
(50-354/86-45-08)
Procedures [(RP)ST-003(Q)] for the operation of the NPV and SPV skids may need provisions for the locally controlled collection of a brief grab sample in the idle sample positio Procedures [RP-ST-004(Q)] for the operation of the FRVSV skid may need to be modified so as to limit the time (and therefore
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the collected activity) of the first grab sample that is ob-tained following the onset of post-accident concentrations of radiog?ses, so as to facilitate its removal from the shield and its transpor *
Spare loaded particulate filter and iodine cannister holders may need to be provided, so that the' sample holder does not have to be unloaded on the spot by hand when grab (or other high acti-vity) samples are removed from the shielded mid-htgh range sam-ple shield in order to install a fresh particulate filters and iodine canniste *
The procedures for the operation of the NPV and SPV skids may need to include the requirement that samples from them be purged prior to their transpor *
The local readout for the FRVSV may need to be moved to a lo-cation that would be visible to personnel performing a sample chargeout, or a supplementary visible readout provide . _ . . . - . _ . - . . . . - - - .
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7.0 Containment High-Range Monitor, Item II.F.1-3 7.1 Position NUREG-0737, Item II.F.1-3, requires the installation of two in-con-tainment radiation monitors with a maximum range of I rad /hr to 10'
rad /hr (beta and gamma) or alternatively 1 R/hr to 10' R/hr (gamma <
only). The monitors shall be physically separated to view a large portion of containment and developed and qualified to function in an accident environment. The monitors are also requiced to have an energy response as specified in NUREG-0737, Table II.F.1- .2 Documents Reviewed .
The implementation, adequacy, and status of the installed in-contain-
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ment high range monitors were reviewed against the criteria set forth in Section 3.0 of this report and in regard to interviews with cogni-zant licensee personnel, licensee letters, station procedures, as-built prints and drawings as l'isted in Attachment 5 to this Inspect-ion Report, and by direct observatio .3 System Description The licensee has installed two General Atomic RD-23 ion chambers (gamma only). The detectors (RE 4825A and B) are located 180 apart at the 145 foot elevation of the drywell. The detectors are powered via separate vital instrument power supplies. The readings from the detectors are not used to determine extent of core damage but are used to modify emergency levels. Detector readout is available in the Control Roo .4 Findings Within the scope of the review, the following itens were reviewed and verified to conform with NUREG-0737:
detector location
electrical separation
range and energy response
vendor type calibration
onsite calibration
- redundancy The establishment and implementation of Technical Specification re-quired surveillance procedures was also verifie Within the scope of this review, the following matters items were identified which the licensee indicated would be reviewed for pos-sible clarification or improvement: (50-354/86-45-09)
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The Containment High Range Monitor is not described in Control Room Operator or Radiation Protection Alarm response procedure *
The monitor surveillance procedure (IC-CC-SP-041(Q)) does not verify alarm annunciation (signal transfer from the RM-80 to the RM-11).
Documentation was not available to substantiate environmental qualification of the installed cable / connector configuratio .0 Improved In-Plant Iodine Instrumentation Under Accident Conditions, Item III.D. .1 Position NUREG-0737, Item III.D.3.3, requires that each licensee provide equip-ment and associated training and procedures for accurately determin-ing the airborne iodine concentration in areas within the facility where plant personnel may be present during an acciden .2 Review Criteria The implementation, adequacy and status of the licensee's in plant iodine monitoring under accident conditions were reviewed against the criteria listed in Section 3.0 and in regard to the documents ident-ified in Attachment 6 to this Inspection Report. The licensee's per-formance relative to these criteria was determined by:
Interviews with cognizant licensee personnel; Review of applicable operational and emergency plan procedures; Review of applicable lesson plans and training records;
Discussions of methodology and implementation with radiation protection technician; and
Verification of equipment availability and storag .3 Description of Methodology and Capabilities Three methods will be used to determine the airborne concentration of iodine within the facility. Radiation exposure dose rates emanat-ing from silver zeolite cartridges dictate the method to be used. If exposure rates from the cartridge are low, a Nal or Ge-Li system will be used for determination of iodine collected. Higher dose rates require use of a surycy meter and conversion factors to determine iodine collected. The air samples are collected via a 9.5 liter evacuated marinelli or a low volume sample pum .
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8.4 Findings Within the scope of this review, the following was identified:
The licensee's program to sample and analyze radioiodine against a background of noble gases was found to be generally acceptable. How-ever, the licensee indicated the following items would be reviewed for possible clarification or improvement: (50-354/86-45-10)
Evaluate the acceptability of using a silver zeolite cartridge
" spiked" with a radioactive button source for use in determing the efficiency of SAM-2 single channel analyzers. This geometry is not consistent with iodine deposition to be expected ( face loaded, uniform iodine deposition).
Provide periodic " hands on" training with emergency sampling equipment. Currently, training consists of procedure walk throughs with some hands on training during emergency exercise Some technicians were uncertain as to how to set-up and operate emergency sampling equipmen *
Consider upgrading Procedure EP-IV-112 and/or EP-1V-113 to add-ress the following. The licensee indicated the procedures would be reviewed for possible clarification or upgrade:
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describe location of dedicated sample pumps (some techni-cians were uncertain of location)
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provide minimum sample volume for low volume air samples (as appropriate)
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describe sample " turnover" requirements to ensure Radio-active Sample Coordinator receives samples
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clarify / describe sample documentation requirements
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define an " abort point"
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identify nitrogen purge location for purging sample cart-ridges
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define / provide guidance for the terms " habitable / access-ible"
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provide maximum allowable background to count / analyze sam-ple cartridges
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describe survey meter geometry used for estimating iodine deposited on charcoal cartridges (i.e. open/ closed window)
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correct dose rate criteria for selecting method to analyze
, sample based on size of source and geometry
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establish equation for determination of iodine airborne activity when using the SAM-2
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identify location of Ba-133 check sources for verifying efficiency of SAM-2. Sources were not contained in Emer-gency Lockers
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identify minimum acceptable vacuum required for use of liter evacuated marinellis
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describe geometry of cartridge when using SAM-2
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identify acceptable sample storage locations
Ensure all appropriate controlled procedures are located in the Technical Support Center. Procedure RP-SA-ZZ-002 was missing from the controlled procedures
Establish procedure RP-TI.ZZ-031. This procedure is referenced in other emergency procedures. The procedure is to replace a chemistry sample analysis procedur Consider dedicating additional SAM-2s to Emergency Locker Only one SAM-2 is dedicated for the TSC, OSC and in plant act-ivitie .0 Quality Assurance & Design Review 9.1 General As part of the inspection effort a review was performed to verify and validate the adequacy of the licensce's design and quality assurance program for the installation of the Post Accident Sampling Syste .2 Documents Reviewed The inspector reviewed pertinent work and quality assurance records for the design, construction and installation of the Post-Accident Sampling System to ascertain whether the records reflect work accomp-lishments consistent with NRC requirements and criteria contained in appropriate documents. The documents reviewed are presented in Attachment 7 to this repor .3 Installation In addition to reviewing the above referenced documents, the inspec-tor verified the PASS installed configuration by determining the following:
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The PASS provides a centralized location for sampling of the reactor coolant and containment atmospher *
The system consists of a prefabricated sampling station with lead brick shielding, control panel, sample cooler, sample cask and positione *
Sampling points are provided to obtain air samples from the suppression chamber, drywell and the reactor building atmos-pher Reactor coolant samples are obtained from sample lines to both RHR loops, the suppression pool and the jet pump pressure in-strumentation line *
The process sample lines connected to the reactor coolant pres-sure boundary (RCPB) through the first isolation valve outside the containment are designed to Category I requirement *
All sample lines beyond the piping-to-tubing reducers conform to quality group D and meet the requirements of ANSI B31.1 Power Piping Cod *
All isolation valves are located in the seismic category 1 por-tion of the sample lin .4 Finding Generally, all QA and design control requirements and procedures were adequately performed. Design changes, nonconformances and rework were .dequately controlled and documented. Records are easily re-trievable, current and signed by authorized cognizant personnel. The design and quality assurance program for the PASS installation was considered. The licensee indicated the following matters would be reviewed for possible clarification or improvenent (50-354/86-45-11):
Power to the PASS isolation valves, control valves, sample station control panel and isolation valve control panel comes from a non-1E battery backed power source. However, in review-ing drawing No. E-0018 the inspector noted that power to an elevator (31-04RM) used in transporting PASS samples to the chemistry laboratory on elevation 124 came from the emergency diesel generator bus through motor control center MCC-00B47 Note 10 of drawing E-0018 indicated that all non-Q loads (included elevator 31-04RM) were dropped from the diesel bus on loss of offsite power /LOCA. The drawing shows a push-button switch in the circuit that would require operator action to place the elevator back on lin Procedure No. CH-E0.SH-007(Q)
for transporting the PASS sample does not address operator act-ion required to ensure availability of elevator 31-04RM in trans-porting of PASS sample to the chemistry laboratory at elevation 12 _ qq,,t s v --
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The PASS Operatiori and Maintenance Instruction, Manual dated I, '
November, 198? does not reflect the as-built ccDfiguratio General Elect ic, the manufacturer, has made alaumber cf instru- e
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menti loop wir'inicorrections and component' changesTrer,laccrent dating back to September, V983 which have not yeri be'en sincor ' N porated in the manual or -dkplicable P&IC's ud conn).ctio'n draw-ing N
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Discussionswiththelicenseeandrepresentativesfrombeneral
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Electric indicate all changes / revisions will be incorporated in an updated manual by December 198 ,
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Calibration stickers on PASS instrument components indicate a 3 i ,,
year calibration schedule. General Electric's Operation and Instruction M.i.ual (bEK-83344, pg 4-12 for FApS, requires
- quarterly readings and yearly operation of the sarnplina , syste FSAR section 9:3.2.3.ih item J.6 states: " equipment usernfpr post accident samplinfi' analysis will be. calibrated or tested approximately every six months." At tb6 present time th4 lic-~ 1 ensee's program requires that sampling be done every six' months ,
with instrument calibration every 18 month The licensee indicyted he would clarify Ms calf bpatjop; reauire-ments if needed., "l '
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C During the samp1fng exercise for PASS on September 24, IR6 the
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inspector noted'that licensee personnel cor,ds. ting PASS; test were using an eight inch cre cent wr'e'nch to turn and hob the '
's control switch no. HC-652. Personnel inaicated excessivb spring 't tension in the switch made ithif '
fficult to rotate and hold for the required three minutei ,
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17 s-Discussions with the.lic5nsee indichte that an adaptor will be ,G i provided for the edditional leverage requifed to operate switch' '-
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no. HC-65 *
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10.0 Licensee Commitmeng \ '
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the licensee provided the follow lr.g cousntments to'addres; the deficiencie !, . - '
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A chemistry department directive was issad to prmildb more specific guidance relative Post Accident to certifying Sampling System. personnelp ,T ' 'quelifigd la operate the ' ' -
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The Post Accident Sampling System procedure wiI? be revised by . 3 September 29, 1986 to provide clearer guidance for dissc!yed gas -
sampling and resolve problems encountered during dissolved gas '
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Prior tc reactor start-up, at least one chemistry technician or technical assistant per shift will be fully qualified in revised Post Accident Sampling Procedures. By October 10, 1986, all chemistry
- tech ~nicians and technical assistants will be qualified on the revised
- l procedure ,
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Self contained breathing apparatus (SCBA) training for chemistry
, technicians and technical assistants will be expedite Should SCBA
use be required prior to completion of the training, radiation pro-
, tection personnel will provide on-the-job training. All SCBA train-ing of chemistry technicians and technical assistants will be cca.-
s pleted by November 1, 198 ;
's Radiation Protection Technicians will receive hands-on-tr.aining on
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in plant iodine sampling equipment on an expedited basis. At least
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one radiation protection technician per shift will be trained prior to start-u 't 11.0 Exit Meeting L Q'
The inspectors met with licensee personnel (denoted in section 1 of the
report) on September 26, 1986. The purpose, scope and findings of this ( 4 ,,
inspection were discusse (-* w
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Attachment 1 Individuals Contacted During Inspection 50-354/86-45 Public Service Electric and Gas Company R. S. Salvesen, General Manager J. R. Lovell, Radiation Protection and Chemistry Manager P. M. Krighva, Assistant to General Manager D. J. Vito, Senior Licensing Engineer M. F. Metcalf, Principal Engineer, QA J. Clancy, Principal Health Physicist R. T. Griffith Sr., Principal Qa Engineer C. G. Atkinson, Engineer T. J. Murpny Engineer D. P. Gauguly, Senior Staff Engineer - EQ G. M. Suey, Chemistry Supervisor M. W. Meltzer, Training Coordinator J. Hawrylck, Engineer L. M. Piccirelli, QA Engineer G. Morrill, Radiation Protection Supervisor, Effluent and Shipping A. Garrison, Training Supervisor - Chemistry D. Parkes, Training Supervisor - Radiation Protection D. Kasner, Training Specialist J. Morgan, Sponser Engineer
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Contractors J. E. Jones, Watts-Bar Jones, In E. Skeehan, GE S. Lee, GE _ _ . . - - _ _ _ - _ . _ . _ _ _ .
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Attachment 2 to Inspection Report 50-354/86-45 Documentation for NUREG-0737. II. Hope Creek Generating Station Procedures-CH-EO.SH-001(Q) Post Accident Sample Panel Operation-CH-E0.SH-004(Q) Post Accident Sample Analysis-CH-TI.ZZ-011(Q) Estimation of Reactor Core Damage Under Accident Conditions-CH-EO.SH-005(Q) Post Accident Sample Dilution-CH-E0.SH-007(Q) Transporting Post Accident Sampling System Samples-CH-GP.ZZ-003(Z) General Safety and Laboratory Practices-SA-AP.ZZ-046(Q) Radiological Access Control Program-SA-AP.ZZ-024(Q) Radiological Protection Program-SA-AP.ZZ-014(Q) Station Personnel Qualification and Training-CH-CA.ZZ-039(0) Anion and Cations by Ion Chromatography-CH-CA.ZZ-0021(Q) pH-CH-RC.ZZ-007(Q) Gamma Spectroscopy Sample Counting-CH-CA.ZZ-025(q) Boron by Specific Ion Electrode-CH-CA.ZZ-040(Q) Hydrogen, Nitrogen and Oxygen by Gas Chromatography-CH-DC.ZZ-002(Q) Calibration of Hydrogen and Oxygen Monitors Hope Creek Generating Station Drawings-M-38-0 Post Accident Sampling System P& ids General Electric Co. Documents
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PASS Suggested Laboratory Practices and Suggested Analytical Procedures, EAF-90-13
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Procedure for Monitoring Total Dissolved Gas Concentration
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Volume 1 of GE PASS Manual, GEK-83344
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Attachment 3 g ana rlson of Chemica l and Radiochemical Test Results Pa rameter Concent ra t ion Meas. Con D i ffe rence License Commitment in FSAR Table 9.3-7 Boron ( Standa rd ) 985 i 10 ppm 1000 ppm + 15 ppe i 50 ppa over the range 50-1007 ppe Diluted 1:1000 2980 1 50 ppm 2700 ppm -120 ppa -----
4870 1 60 ppe 4900 ppe + 30 ppe -----
Chloride 24.1 1 3.1 ppm * 23.0 ppa - 1.1 ppe i 10% over the range 0.5 to 20 ppe ( Standa rd) 37.4 1 1.2 ppm * 38.9 ppe + 1.5 ppe
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i 50 ppb <500 ppb Chloride 24.1 1 3.1 pp .4 i .7 ppa ( Standa rd ) 37.4 1 1.2 pps 42 1 2 + 4.6 ppe (Analyzed by 80.5 1 2.2 ppm 64 1 3 -16.5 ppa effsite lab:
BaM, Lynchburg, VA)
pH*** .9 - .3 pH units over the range 5 to 9, (cctual samples) .9 - i O.5 pH units <5 or >9 Gemma Isotopic (3.210.3)E-5 mci /ml (3.910.4)E-5 mci /ml + 0.7 mci /mi i2 C2-58 (cctual samples)
NOTE: Normal Boron calibration curve renses from 50-10,000 ppa with 100 to 1 dilution factor
- diluted 1:1000
- diluted 1:2000
- pH of 6.3 was measured on RHR inline sonitor at 0545 hrs 9/24/86. Located at normal sample panel sink
- pH of 6.1 was measured on RHR sample taken at 1325 hrs on 9/24/86 from normal sample panel sink
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Attachment 4 Documentation for NUREG 0737, II. F.1-1 & 2 Public Service Electric & Gas Company - Hope Creek 50-354 Licensee Operating Procedures !
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RP-SD-SP-002(Q) " Operation of the FBVSV Skid" Rev. O, dated March 10, 1986
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RP-SR-002-CH "FRVS Rad Monitor Source Check-Monthly", Rev. 5,
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RP-TI-SP-005(Q) " Operation of the North and South Plant Vent Skids, Rev. 1, dated March 10, 1986
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RP-TI-SA-004(Q) " South Plant Vent Monitor Source Check-Monthly", Rev. 5,
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RP-ST-SP005(Q) North Plant Rad Monitors Source Check-Monthly, Ray. 5,
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RP-TI-ZZ-024(Q) Gaseous Effluent Permit Generation (Rev.1), dated June 19, 1986
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RP-AR-SP-001(Q) Radiation Monitoring Gaseous Monitor Alarm Response (Rev. 1), dated June 19, 1986 l
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RP-TI-22-025(Q) Semi-Annual Report Generation, Rev. O, dated June 24, 198 RP-TI-SP-003(Q) Operation of the North and South Plant Vent Skids, Rev. 1, dated April 8, 1986
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RP-TE-SP-006(Q) "WRGM Mid/Hi Range Monitor Energy Range Determination, Rev. O,
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l RP-ST-ZZ-004(Q) Gaseous Effluent Surveillance Rev. RP-AP-SP-001(Q) Control of Radiation Monitoring System Setpoints, Rev. O, dated March 7, 1986
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RP-TI-SP-001(Q) Operation of the General Atomics RM-23, Rev. 1, dated March 25, 1986
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RP-TI-SP-011-(Q) RM-11 Operating Procedures, Rev. O, dated August 13, 1986
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CH-RC-007(Q) " Gamma Spectroscopy Sample Countings, Rev. 1, March 14, 1986
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RP-SA-ZZ-002(Q) " Airborne Radioactivity Analysis"
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RP-TI-SP-001 " Operation of the General Atomic RM-23" Rev. 1, dated March 25, 1986
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RP-AP-ZZ-014(Q) " Training and Qualification Program", Re TP-607-HC-967-1.38.1-01 " Instrument and Control (Computers) Technician '
Training Program, Site Specific Training, GA Technologies Digital Radiation Monitoring System", Rev. 1, Draft, dated July 1, 1986 Sorrento Electronics Manuals GA Technologies Manuals
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E-115-865 (Rev. 3) Wide Range Monitor Equipment Manual, March 1986
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E-115-1318 Maintenance
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E-255-1318 P.I.G. Radiation Monitor System, dated December 198 E-155-1067
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E-115-1097 RM-11 Operations Guide, dated April 1982
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E-115-819 Radiation Monitoring System Communications Software Document, Rev. 3, dated March 198 E-255-961 Calibration Report RD-72 Wide Range Gas Monitor High and Mid-Range Detectors, Rev. 7, dated August 198 E-115-647 Calibration Report for Model 52 Off-Line Beta Detectors, Re , dated October 198 E-115-1304 Digital High-Range Radiation Monitor System for Hope Creek Generating Station Unit 1 Licensee Calibration Procedures
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RTF-HPP 86-004 Efficiency Determination, dated August 11, 1986
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IC-SC-SP-001(Q) Sensor Calibration, Filtration, Recirculation Vent Process Flow, Rev. O, dated April 11, 198 IC-SC-SP-002(Q) Sensor Calibration, Filtration, Recirculation and l Ventilation Sample Flow System, Rev. O, dated April 4, 198 IC-SC-SP-004(Q) Senor Calibration, South Plant Vent Flow Rate Monitor, Rev. 1, dated May 23, 198 IC-SC-SP-005(Q) Sensor Calibration, South Plant Vent Flow Rate Monitor, Rev. O, dated April 11, 198 _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
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IC-SC-SP-006(Q) Sensor Calibration, North Plant Vent Process Flow Rev. O, dated April 11, 198 IC-SC-SP-007(Q) Sensor Calibration, North Plant Vent Sample Flow System, Rev. O, dated April 8, 198 IC-SC-SP-008(Q) Sensor Calibration, Filtration Recirculation Vent Noble Gas, Rev. 1, dated May 9, 198 IC-SC-SP-009(Q) Sensor Calibration, Filtration Recirculation Vent Noble Gas, Rev. 1 dated May 12, 198 IC-SC-SP-010(Q) Sensor Calibration, Filtration Recirculation Vent Noble Gas, Rev. I dated May 13, 198 IC-SC-SP-014(Q) Sensor Calibration, South Plant Vent Noble Gas, Rev. 1, dated April 25, 1986
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IC-SC-SP-015(Q) Seasor Calibration, South Plant Vent High Range Rev. 1, dated May 9, 1986
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IC-SC-SP-020(Q) Sensor Calibration, North Plant Vent High Range Rev. 1, dated May 15, 1986
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IC-SC-SP-021(Q) Sensor Calibration. North Plant Vent High Range Rev. 1, dated May 13, 1986
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IC-SC-SP-022(Q) Sensor Calibration, North Plant Vent Mid Range Rev. 1, dated May 13, 1986
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IC-SC-SP-027(Q) Sensor Calibration, Drywell Atmosphere Post Accident Rev.1, dated April 30, 1986
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IC-CC-SP-015(Q) Channel Calibration, North Plant Vent (WRGM) Rev. O, dated April 11, 198 IC-CC-SP-021Q) Channel Calibration, South Plant Vent (WRGM) Rev. 1, dated July 29, 1986.
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IC-CC-SP-031(Q) Channel Calibration, Filtration Recirculation Vent, Rev. O, dated April 19, 198 IC-CC-SP-041(Q) Channel Calibration, Drywell Atmosphere Post Accident (DAPA) dated April 23, 198 IC-CC-SP-042(Q) Channel Calibration, Drywell Atmosphere Post Accident (DAPA), Rev. O, dated February 25, 1986.
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Licensee Training Procedures
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TP-607-H6-967-1.38.1-01 " Site Specific Training GA Technologies Digital Radiation Monitoring System, Rev. 1, dated July 1, 1986
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RP-TI-SP-001(Q) " Operation of the General Atomic RM-231, Rev. 1, dated March 26, 1986
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RP-TI-SP-003(Q) " Operation of the North and South Plant Vent Skids",
Rev. 1, dated April 8, 1986
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RP-S0-SP-002(Q) " Operation of the FRVSV Skid", Rev. O, dated March 10, 1986 Licensee Emergency Procedures
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EP-105 " Evaluation of RMS Data from High Range Channel RP-48118 (DAPA),
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EP-106 " Computerized Effluent Dose Assessment", Rev. 2, dated February 1, 1986
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EP-107 " Manual Effluent Dose Assessment", Rev. 2, dated February 2, 1986
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EP-IV-104 " Emergency Exposure Authorization and Emergency Task Alarm Review", Rev. 1, dated October 17, 1986
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EP-IV-106 " Computerized Effluent Dose Calculations", Rev. 2, dated February 1, 1986
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EP-107 " Manual Effluent Dose Assessment", Rev. 2, dated February 2,1986
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EP-IV-104 " Emergency Exposure Authorization and Emergency Task Alarm Review", Rev. 1, dated October 17, 1986
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EP-IV-106 " Computerized Effluent Dose Calculations". Rev. 2. dated February 1, 1986
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EP-IV-111 " Coordination and Storage of Post-Accident Samples", Rev. O, dated May 10, 1986
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EP-IV-113 "Radioiodine and Noble Gas Activity Sample Analysis During An Emergency", Rev. 1, dated October 15, 1985 Licensee Chemistry Procedures
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CH-RC-ZZ-007(Q) "Ganima Spectroscopy Sample Counting, Rev. I f
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Licensee Operating Procedures
--0P-E0-ZZ-004(Q) " Radioactivity Release", Rev. 0 Licensee Pre-operational Test Procedures
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PTP-SP-2 " Class II Radiation Monitors to RM-11 Computer, June 24, 1986
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PTP-SP-3 "Non IE Process Radiation Monitors, Cooler Condensate Monitor and Door Monitoring to RM-11 Computer, Pre-Operational Test
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PTP-SP-210 " Class II Radiation Monitors to RM-11 Computer, August 19, 1986 l
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DTP-SP-004 North Plant Ventilation and South Plant Ventilation Radiation Monitoring System (NPV & SPV RMS) Rev. O, '
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DTP-SP-0005 " Filtration Recirculation and Ventilation System Vent Radiation Monitoring System (FRVSV RM1), Rev. O, dated February 8,1986 Licensee Drawings - PSE&G (Job 10855, Bechtel)
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0366-9026, " Acceptance Test Report WRGM System - FRVSV" #1
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0366-9010, " Transfer Calibration Report, Wide Range Gas Monitor", TR #1, 2, 3
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M-26-1 Sheet 1, " Radiological Monitoring System, Rev.1, dated February 7,1986
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M-26-1 Sheet 2, " Radiological Monitoring System Rev. 5, dated July 18, 1986
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E-0012 Sheet 3, " Single Line Relay Diagrams, 120V AG Instrumentation and Misc Systems", Rev. 8, August 25, 1986 L
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E-12280 Sheet 3, " Cable Block Diagram, Air Radiation Monitoring Systems, North and South Plant Vent & FBVSV System, Rev.1, dated April 10, 1986
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, Attachment 5 to Inspection Report 50-354/86-45 Documentation for NUREG-0737 Item II.F.1-3 General Atomic
Transfer Calibration Reports - Ion Chamber Area Monitors RD2A, RD2BA and RD23 (DAPA Channel A and B) dated July 10, 1985
Energy Response Test and Dose Rate Calibration of Model RD-23 High-Range Radiation Monitor, dated August 1, 1985
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High Range Radiation Monitor - Equipment Manual, dated January 1986 l
Acceptance Test Report - DADA Channel B, A dated July 16, 1985
Transfer Calibration Procedure - Ion Chamber Area Monitors - RD2A, RD2B &
RD23, dated July 16, 1985 Hope Creek Procedure IC-CC.SP-042(Q), Channel Calibration Process Radiation Monitoring - Non Divisional Monitor RY-4825B - Drywell Atmosphere Post-Accident (DAPA), dated February 25, 1986
Procedure IC-SC-027(Q), Sensor Calibration Process Radiation Monitoring -
Non Divisional Channel B R-4825B Drywell Atmosphere Post-Accident, dated September 3, 1986 Procedure IC-SC.SP-026(Q), Sensor Calibration Process Radiation Monitoring - Non Divisional Channel A R-4825A Drywell Atn.osphere Post-Accident
Procedure IC-CC.SP-041(Q), Channel Calibration Process Radiation Monitoring -Non Divisional Monitor RY-4825A Drywell Atmosphere Post-Accident (DAPA), dated April 23, 1986
Procedure OP-ST.SH-001(Q), Accident Monitoring Instrumentation Channel Check - Monthly, dated July 16, 1986
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t Attachment 6 to Inspection Report 50-354/86-45 Documentation for NUREG'-0737 Item IIID. Hope Creek Generating Station Procedures
EP IV-113, "Radioiodine and Noble Gas Activity Sample Analysis During an Emergency"
EP IV-112, " Emergency Air Sampling"
Memorandum: "
Improved Inplant Iodine Instrumentation Under Accident Condition", dated May 3, 1985 (Manager, RPS to Hope Creek, RPM)
RP-SA.ZZ.002(Q), Rev 3, " Airborne Radioactivity Analyses, September 16, 1986
RP-AP.ZZ-014(Q), Rev 1, " Training and Qualification Program", March 21, 1986
RP-TI.ZZ-018(Q), Rev 3, " Routine Operation of Counter Scalers", dated September 16, 198 M12-ICI-204, Rev 0, " Calibration of the SAM-2/RD-19/RD-22," July 27, 1985 0 .M12-ICI-107, Rev 0, " Calibration of R0-A and R0-2A Ion Chamber"
M12-ICI-303, Rev 0, " Calibration of Rodeco H-809(V) Air Samplers" e
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Attachment 7 Documents Associated with Design, Construction and Installation of the Post-Accident Sampling System
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System Design Specification No. 10855-D5.16
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Operation and Maintenance Instruction Manual No. GEK-83344 (Vendor Print No. 10855-N1-D24-5)
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General Electric Design Specification No. 22A7465 rev 1
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NUREG-0737 Clarification of TMI Action Plan Requirements
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ANSI Stanaard 83.1 Power Piping Code
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ASME Boiler and Pressure Vessel Code,Section III Class 1, 2 and 3
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Regulator Guide 1.97 - Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident ,
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ANSI /ANS 4-5-1980 Criteria for Accident Monitoring Functions in Light-Water Cooled Reactors
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NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and Short Teru Recommendation Drawing No. FSK-JD-7201-1-005-1,-2, -3 and -4 line from suppression poo Drawing No. FSK-JD-7201-1-013-1 & FSK-JD-7201-016-1 flow line from piping statio Drawing No. FSK-JD-7201-1-009-1 return line to suppression pool.
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Drawing No. FSK-JD-7201-1-003-1, -2, -3, -4, -5, -6 & -7 line from drywell atmosphere
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Drawing No. FSK-JD-7200-1-008-2, -3 & -4 line from RHR loo Drawing No. 0-P-RC-201 rev. 4, sh I &2 Torus sample Ifne P-22 Drawing No. 0-P-RC-203 rev. 4, drywell sample point from penetration JT Drawing No. 0-P-RC-207 rev. 1, sample line from jet pum Valve Qualification Report No. QR526-5683-6 ref. I and QR52600-515 ref. Equipment Evaluation Summary Sheets No. J603-SV-002 rev 4.
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Procedure No. MD-GP-ZZ-033(Q) rev. Purchase Order No.10855-J-603(Q)
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Design Specification No.10855-J-603(Q)
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Quality Control Inspection Report Nos. 0-P-RC-201-1-P-1.10, 0-P-RC-201-1A-P-1.10 and 0-P-RC-201-2-P-1.10
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Quality Control File No. 3013-M38-0-1
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Weld /Non-Destructive Specification Nos. P-202(Q) rev. 10, P-500(Q) re and P-205(Q) rev. Dye Penetrant Etamination Procedure No. PT-HC-77-9-12-3 rev. 3.
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Weld Procedure No. P8-T-AG rev. ,
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