IR 05000423/1986027

Safety Insp Rept 50-423/86-27 on 860818-22.No Violations or Deviations Identified.Major Areas Inspected:Status & Implementation of NUREG-0737 Items,Including post-accident Sampling of Reactor Coolant & Containment Atmosphere

ML20215H057
Person / Time
Site:	Millstone
Issue date:	10/09/1986
From:	Cioffi J, Mark Miller, Musolino S, Pasciak W, Weadock T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20215H054	List: IR 05000423/1986027 ML20215H051
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-2.F.1, TASK-3.D.3.3, TASK-TM 50-423-86-27, GL-82-05, GL-82-5, NUDOCS 8610210398
Download: ML20215H057 (22)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Report No. 86-27 Docket N License No. NPF-49 Category C Licensee: Northeast Nuclear Energy Company P. O. Box 270 Hartford, Connecticut 06141-0270 j

Facility Name: Millstone Point Unit 3 Inspection At: Waterford, Connecticut Inspection Conducted: August 18 - 22, 1986 Inspectors: 75/(thL 1; fins -

9/atf/JC M. MM1er, Radtstion Specialist 'dat

% ask J.yioffi,Radiaf!f6nSpecialist bla

'dats OM &

T. padock, Radiatifon Speciap st 1/u/n date '

p NYb Y 29 S. Musol , ontractor BNL Bate Approved by: DW - O lob fIC, W. Pasciak, Chief, Effluents Radiation @tel Protectior. Section Inspection Summary: Inspection conducted on August 18 - 22, 1986 (Inspection Report No. 50-423/86 ~2 Areas InspecteA: Special, announced safety inspection of the licensee's implementation end status of the following task actions identified in NUREG-0737: Post-accident sampling of reactor coolant and containment atmosphere; increased range o' radiation monitors; post-accident effluent monitoring; containment radiat'on monitoring; and in plant radiciodine measurement Results: -There were no violations or deviations identified during this revie However, several areas were identified which require improvements. For three of the items identified in this report, as 86-27-01 (paragraph 4.3.1),86-27-04 (giaragraph 6.4), and 86-27-06 (paragraph 7.3), you are requested to respond within 30 days of receipt of this report on corrective actions taken or planne ~

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DETAILS Personnel Contacted 1.1 Licensee Personnel

• W. Romberg, Station Superintendent, Millstone Point

• J. Kangley, Radiological Services Supervisor, Millstone Point

• J. Waters, Chemistry Supervisor, Millstone Point

• K. Burton, Millstone Unit 3 Operations Supervisor

• R. Sachatello, Millstone Unit 3 Health Physics Supervisor

• J. Harris, Millstone Unit 3 Engineering

• R. Rothgeb, Millstone Unit 3 Maintenance

• E. Peterson, Millstone Unit 3 Instrument & Controls

• Tortora, Chemist, Millstone Unit 3

• Closius, QA/QC Supervisor, Millstone Point

• T. Burns, Assistant Chemistry Supervisor

• F. Lukaszek, NSEDG, NUSCO T. Rogers, Department Planner R. Beckman, I&C Specialist W. Buck, Emergency Planning T. Burns, Chemistry Supervisor T. Cleary, Engineer K. Cox, I&C Specialist R. Crandall, Radiological Engineer R. Enoch, I&C Supervisor R. Olsen, I&C NUSCO F. Perry, Assistant Radiation Protection Supervisor M. Samek, I&C Engineer D. Scace, Engineer 1.2 NRC Personnel

• Cioffi, Radiation Specialist, RI

• Conner, Acting Resident Inspector, RI

• k. Gregg, Reactor Engineer, RI

• M. Miller, Radiation Specialist, RI

• S. Musolino, Health Physicist, BNL

• T. Weadock, Radiation Specialist, RI

• Denotes attendance at the Exit Interview conducted on August 22, 198 Other members of the licensee's staff were also contacted and/or partici-pated in exercises of post-accident and effluent monitoring systems during the inspectio ._ - . . - - - -. ._. --- .-- _. --

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3 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 In-Plant Iodine Instrumentation Under Accident Conditions 3. TMI Action Plan Generic Criteria and Commitments The licensee's implementation of the task actions specified in Section were reviewed against criteria and commitments contained in the following documents:

NUREG-0737, Clarification of TMI Action Plan Requirements, dated November 198 *

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, 198 NUREG-0578, TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations, dated July 197 *

Letter from Darrell G. Eisenhut, Acting Director, Division of Opera-ting Reactors, NRC, to all Operating Power Plants, dated October 30, 197 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 II.F.1, Attachments 1, 2 and 3, NUREG-0737" dated August 16, 198 *

Safety Evaluation Report related to tha operation of Millstone Nuclear Power Station, Unit No. 3 Docket No. 50-423, dated July 198 *

Supplementary Safety Evaluation Report related to the operation of Millstone Nuclear Power Station, Unit No. 3, Docket No. 50-423, Supplement No. 3, dated November 198 . _ . - .- - _

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Supplementary Safety Evaluation Report related to the operation of Millstone Nuclear Power Station, Unit No. 3, Docket No. 50-423, Supplement No. 4, dated November 198 *

Regulatory Guide'1.4, " Assumptions Used for Evaluating Radiological Consequences of a Loss of Coolant Accident.for Pressurized Water Reactors".

Regulatory Guide 1.97, Revision 3, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an 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".

4. Post-Accident Sampling System, Item II. .1 Position NUREG-0737, Item II.B.3, specifies that licensees shall have the capability to promptly collect, handle, and analyze post-accident samples which are representative of conditions existing in the reactor coolant and containment atmosphere. Specific criteria are denoted in commitments to the NRC relative to the specifications J

contained in NUREG-073 Documents Reviewed The implementation, adequacy and status of the licensee's post-accident sampling and monitoring 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 1.A 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-

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System Description and Capability

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The licensee has installed a Post-Accident Sampling System which is a standard General Dynamics design. It has the capability to obtain pressurized and unpressurized atmosphere and coolant samples. Liquid samples can be taken from the reactor coolant hot legs and from the containment sumps. (Atmosphere samples can be obtained from the hydrogen recombiner supply lines.) The PASS sampling cabinet and control panels are located in rooms just outside the hydrogen recombiner _ ._ -

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4.2 Findings Within the scope of the review, the following items were identified:

4. Performance Test Grab samples of reactor coolant and of the containment atmosphere were obtained in a performance test for this inspection on August 17, 1986. During the test, licensee personnel verified the integrated ability to collect and analyze samples within the constraints of NUREG-0737, II.B.3. However, equipment malfunctions were observed that .

caused sampling delay times, liquid in the dissolved gas sample and questionable pressure measurements, which were necessary for calculating total dissolved gases. The following observations were noted:

Leakage in the dissolved gas sampling line; ,

Flushing pump tripped five times during performance test;

Pressure indicator was slow to respond and did not stabilize;

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Operation of a valve (V-46) which isolates the con-tainment atmosphere sampling line was hampered for one hour when its reach rod pin was sheare *

PASS instrumentation was on a 2 year calibration frequency which was in excess of the vendor recom-mended frequencie . Sampling - General Observations Surveillance and training programs had been conducted in accordance with commitments documented in the Unit 3 SE Procedures were in place for the required sample pathway However, the following concerns were identified:

The procedure for conduct of the reactor coolin:

sampling operation (EPIP 4224, Revision 2) concained some valve operations out of sequence, included steps that did not specifically list the valves to be repositioned, and did not include the necessary Control Room notification . - - . - - _ - -

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It was noted that the licensee previously identified the valve operation clarifications, sample aliquots to be obtained, and opening of a second ventilation damper. However, these changes were not incorporated into a temporary procedure change at the time of the inspectio The Millstone Unit 3 FSAR, Section 9.3.2.6.2, Process Sampling System Description discussed other sampling pathways, i.e., reactor coolant cold legs, containment -

recirculation and auxiliary building, where PASS samples could be obtained. However, the inspectors noted that these pathways were locked, had not been tested and procedures were not availabl *

The indoor occupancy times documented in Section 12.3.1.3.2 of the FSAR were not conservative and should be revised based on demonstrated PASS sampling time .2.3 Analytical Capability The General Dynamics PASS system provides an in-line anal-ysis capability for the pH of liquid samples and for the establishment of the volume of dissolved gases /kg coolan All other analyses are performed off-line in an analytical laboratory and counting room. The Millstone Unit 3 FSAR did not identify the Unit 3 laboratory facility as acces-sible during an accident. The primary laboratory was identified as the Units 1 and 2. facility with a back-up capability provided by the Haddam Neck Plant. These facilities were previously reviewed and found acceptabl The Safety Evaluation Report documented the Unit 3 labora-tory as the primary analysis facility. Moreover, the Unit 3 facility is fully equipped and capable of performing the required analyses. Therefore, this facility was reviewed against the constraints of NUREG-073 The required chemical analyses utilize the following prin-cipal methods and/or devices:

Boron - Inductively Coupled Plasma

! Chloride - Ion chromatograph Dissolved Gas - Gas Chromatograph Gamma Analysis - Three GeLi detectors and a computerized spectrometer system for the identification and quantification of gamma emitting nuclides.

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Verification of the licensee's ability to fulfill the chemical analytical requirements of NUREG-0737 II.B.3 was made in part by the preparation and analyses of "known" spiked sample In addition to the analysis of the regular samples from the PASS systems, additional verification of the radioactivity analytical requirements was made by the comparison of large volume samples from the PASS systems with those obtained from the " normal" sampling system for the same sampling locations. A summary of the licensee's

performance relative to analytical capability is contained in Attachment I The overall performance of the licensee was found accept-able, with one minor exception:

The Millstone Unit 3 laboratory should be identified as an accessible vital area after/during an accident in Section 12.3.1.3 of the FSAR. If calculations can't support the use of this facility, then related procedures should address the Millstone 1 & 2 lab facilities; and access

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pathways based on the present facility layout should be reviewed in accordance with GDC-19 criteria.

4.3 Recommendations for Improvement 4. PASS Sampling

The equipment problems identified during the perform-ance test must be resolved to ensure PASS sampling capabilit In acdition, the schedule for correcting these problems should be provided as requested during the exit interview (50-423/86-27-01).

4. General Observations Procedure revision to EP-4224 is required to clarify sampling and valve operation *

Review calibration frequencies for PASS Instrumenta-tion to address vendor recommended and standard industry practice *

Review FSAR and submit an amendment change request to reflect actual / planned PASS sampling and analytical capabilitie These items will be reviewed during a subsequent inspection (50-423/86-27-02).

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5. Noble Gas Effluent Monitor, Item II.F.1-1 5.1 Position NUREG-0737, Item II.F.1-1 requires the installation of noble gas monitors with an extended range designed to function during normal operating and accident conditions. The criteria, including the design basis range of monitors for individual release pathways, power supply, calibration and other design considerations are set forth in Table II.F.1-1 of NUREG-073 Documents Reviewed The implementation, adequacy, and status of the licensees monitoring systems were reviewed against the criteria identified in Section and in regard to licensee letters, memoranda, drawings and station procedures as listed in Attachment The licensee's performance relative to this criteria was determined by interviews with the principal persons associated with the design, testing, operation, installation and surveillance of the high-range gas monitoring systems, a reviewsof the associated procedures and documentation, and examination of personnel qualifications and direct observation of the system System Description Two effluent release paths are monitored for high levels of particu-lates, halogens and noble gases from Millstone Unit 3. Each is sampled with a Kaman Model KMG-HRH radiation monitor. In the event of a "high" alarm by the low-range monitor, the sample flow is auto-matically diverted to the high-range monitor pathway, and vice-versa when levels decrease. One decade of range overlap was demonstrate The principal release pathway is the Unit 1 375' stac It includes the radioactive gas waste system, containment vent header, condenser air ejector, other normal radioactive discharges and the supplemental leak collection and release system (SLCRS). The auxiliary and service buildings are normally routed to a Ventilation Ven Under accident conditions flow from this vent is diverted to SLCRS and from it to the Unit 1 stac Ion chamber monitors, Model KDI-F, are installed on the Unit 3 main steam relief lines and the turbine driven auxiliary feed water pump steam exhaust. In the event of a failure of the plant process radi-ation monitoring (RMS) computer, the detectors would be readout locally by an I&C technician. A back-up for the detectors can be provided by a direct measurement with a Teletector Survey Instrumen .

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Surveillance procedures for calibration and functional checks are

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written and have been conducted. Calibrations with Cs-137 sources in I

a fixed geometry are utilized to confirm that the detector response corresponds to the vendor type test.

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5.2 Findings

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Within the scope of this review, the following was identified:

The installed system meets the requirements for high-range noble gas monitoring as contained in NUREG-0737, Attachment II.F.1- .3 Recommendations for Improvement The environmental qualification documentation for the ion chamber used on the main steam relief monitors was not available at the time of the inspection and should be demonstrate *

Written procedures for the implementation of the established alternative methods to readout the main steam relief monitors in the event the RMS computer are unavailable should be pro-vide These items will be reviewed during a future inspection (50-423/.

86-27-03).

6. Sampling and Analyses of Plant Effluents, Item II.F.1-2

6.1 Position NUREG-0/37, 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 that may accompany gaseous effluents following an acciden It must be performable without exceeding the specified GDC-19 dose limits to the individuals invol-ved. The criteria including the design basis shielding envelope, sampling media, sampling considerations, and analysis considerations

are set forth in Table II.F.1- Documents Reviewed The implementat.icn, adequacy and status of the licensee's sampling and analysis system and procedures were reviewed against the criteria identified in Section 3.0 and in regard to licensee letters, memor-anda, drawings and station procedures as listed in Attachment The licensee's performance relative to these criteria was determined by interviewing the principal persons associated with the design, testing, operation, installation and surveillance of the systems for sampling and analysis of high activity radioiodine and particulate i

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effluents, by reviewing associated procedures and documentation, by examining personnel qualifications, and by direct observation of the system .

System Description Isokinetic flow is maintained by the low-range monitor pumping system over a range of 30K-280K CFM in the effluent path. The main sampling lines are 1" diameter. A second pump on the sampling skid delivers 650 cm*/ min to the high-range halogen and particulate filters through about 10 feet of %" lin The sampling skids contain three particulate and iodine sampling channel Each consists of a 2" diameter particulate filter paper followed by a 2.25" by 1" deep charcoal filter cartridge. The collection assembly is located in a shielded sample chamber to protect personnel during filter changes. Each channel has a GM tube to sense radiation buildup on the filter which is set to alarm at 100 mR/hr. Upon alarm, a micro-computer which controls the routing of the flow will automatically switch it to the next filte If no filter changes are made, the system will continue to direct sample flow through the third filte The collection assembly has a quick release and a remote handling tool for the removal and transfer of a potentially high level sampl A portable shield is located in the Chemistry Laboratory and must be carried to the sampling ski However, the licensee had not per-formed calculations to estimate the dose to personnel required to change-out and transport filter cartridges under accident condition The licensee had not developed adequate procedures for the analysis of a filter which may contain high levels of activit .2 Findings Within the scope of this review the following was identified:

A charcoal filter is normally installed in the sampling flowpath of the monitors. However, change over to silver loaded silica gel cartridges will occur at the discretion of the EOF. The licensee had not considered that, given accident level iodine and noble gas concentrations, excessive amounts of noble gas could be retained on the charcoal filter and cause the 100 mR/hr alarm to occur in a minute or less. If this should happen, the sampler could collect two nonrepresentative samples in rapid sequence and then switch over to the third filter. By the time samples were retrieved, the third sample could contain more radioactivity than could be handled within the GDC-19 dose limits or that could be analyzed with the licensee's present analytical capability and procedure _- ._ - . _ . - -_ _ - - - - - - _ _ - _ . . -- - - - , . -. _

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6.3 Acceptability The installed system meets the requirements of NUREG-0737, Attachment II.F.1- .4 Recommendations for Improvement

Improve laboratory procedures for the handling and analysis of high level samples; Demonstrate that plant personnel can collect, replace and trans-port samples within GDC-19 limit *

Incorporate silver zeolite filters in lieu of charcoal prior to switchover to silver loaded silica gel, in order to reduce potential radiation levels due to retained noble gases and to assure a representative sampling techniqu These items will be reviewed during a future' inspection. In addi-tion, the schedule for this item should be provided as requested during the exit interview. (50-423/86-27-04)

7. Containment High-Range Radiation 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 rar.ge of I rad /hr to 10'

rad /hr (beta and gamma) or alternatively 1 R/hr to 107 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 enviranment. The monitors are also required to have an energy response as specified in NUREG-0737, Table II.F. Table II.F.1-3 of NUREG-0737 also outlines specific high-range monitor calibration criteria. These include:

Certification of laboratory calibration of each detector for at least one point per the decade ranges of 10 - 10' R/hr prior to initial us *

Subsequent in-situ calibrations shall include use of a cali-brated radiation source for at least one decade below 10 R/hr; calibration by electronic signal substitution is acceptable for range decades above 10 R/h _

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Documents Reviewed The implementation, adequacy, and status of the installed in-contain-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 listed in Attachment 1.C to this Inspec-tion Report, and by direct observatio System Description The licensee originally installed two Kaman high-range monitors on the inside face of the annulus wall, 51' elevation, to provide post-accident monitoring capability. The inspector determined by review of licensee drawings that this original placement meets the "widely separatea" requirement of Table II.F.1-3. However, upon installation and testing the licensee identified that portions of the insulated cables for this system were defective. The licensee was able to operate one Kaman monitor by employing a short length of acceptable .

cable and re-locating one detector to the 24' elevation of contain-ment. The licensee submitted a special report describing the above to the NRC on January 24, 198 During April,1986, the licensee installed two General Atomics (GA)

high range monitors on the 51' elevation of containment, next to the original location of the Kaman monitors. These monitors were instal-led to meet the containment high-range monitoring requirements of NUREG-0737, independent of the Kaman radiation monitoring syste The licensee indicated that the defective Kaman cabling has been repaired and is currently on-site. During the next outage, this cabling will be run and the relocated Kaman monitor will be restored to its original location on the 51' elevation of containmen Ulti-mately, then, the licensee will exceed the requirements of NUREG-0737 by having four channels for containment area high-range radiation monitoring, provided by two independent system .2 Findings Within the scope of this review, the following was identified:

Vendor calibration data for both the Kaman and GA monitors indicated the detectors met the energy response, range, and type test criteria of NUREG-0737. Vital instrument power supply to the detectors was verifie Procedure EPIP 4212, "Drywell/ Containment Curie Level Estimation", appropriately reflects the multiple monitors presently in containment and appears to provide an adequate method for pro-viding a quick estimate of containment activit . _ - - - .-- - _ - - . . - _ - - . . _ _ _ - _ _ _ _ _ . - _ _ _ ._ ._

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m The inspector reviewed the licensee's calibration procedures for the Unit 3 containment monitors and identified the following deficien-cies:

• Procedure SP 3449H01, used for the Kaman monitor calibration, did not include electronic checks for the 10" - 108 decade * Procedure SP 3449H21, used for the GA monitor calibration, was written such that if a periodic rather than initial calibration was being performed, no electronic checks would be include Additionally, the data sheets included with the procedure did not include areas to record the average detector sensitivity or the results of the one point detector check; both of these steps are required in the body of the procedur The licensee immediately revised the above procedures to address the identified problem During a review of the Millstone Unit 1 and 2 Post-Accident Sampling Systems in 1984 (NRC Report No.~84-09) it was identified that the Unit 2 containment high-range monitor calibration procedure failed to include a single point radiological source calibration (<10R/hr).

In an effort to close the item generated in that report, the current high-range monitor calibration procedures for Units 1 and 2 were reviewed. The following deficiencies were identified:

• Procedure SP4070 is used for the quarterly Unit I high-range monitor calibration. This procedure does not require a single point (<10 R/hr) radiological source calibration; in addition, electronic checks are not performed for all decades of monitor respons ,

Procedure SP4070 is performed for Unit 1 detector change-out, i.e. the initial in-situ calibration of a monitor. This proce-dure does involve exposing the monitor to radiation fields of varying strengths, however, it does not specify that one point per decade between 10 R/hr and 108 R/hr be calibrate * Procedure SP2404AY is used for the Unit 2 containment high-range monitor calibration. This procedure includes exposing the detector to a radioactive source. However, only detector response is tested for, no acceptance criteria for accuracy of monitor response is evaluated. Additionally, an electronic check of the monitor is performed on two decades onl Licensee problems in the area of monitor calibration have not been limited to the above instances. During NRC preoperational inspec-tions of the Unit 3 process and effluent monitors it was noted that the formula for calculating the Radiologically Engineered Calibration Factor (RECF) was incorrectly stated in the calibration procedur i I

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o The RECF is input into the monitor software and converts monitor response to effluent activity. This was corrected during the week of the inspection by the license Licensee Event Report #86-023-00, dated April 10, 1986, describes a calibration problem which occurred with the Unit 3 containment high-range monitors. The licensee identified that these monitors had been miscalibrated and responded in the non-conservative direction by as much as three orders of magnitude. Licensee corrective actions detailed in the report included recalibration of the specific moni-tors in accordance with revised procedures and review of all cali-bration methods for all radiation monitor The inspector stated to the licensee, in light of the recurrent history of problems identified above, that the-current level of attention and technical review directed towards evaluating monitor calibration and calibration procedures does not appear appropriate to assure calibrations are performed adequately to fulfill regulatory

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requirements. In addition, licensee corrective actions in response

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to a calibration concern raised in the 1984 NRC PASS inspection were limite .3 Recommendation for Improvement

The licensee was requested to include in their response to this inspection a description of their intended actions to ensure station radiation monitor calibration procedures were effective-ly reviewed and implemented. This item will remain open and be reviewed in a future inspection (86-27-06).

Installation, calibration, and operability of the Kaman monitors will be reviewed subsequent to the licensee's intended reloca-tion and cable run activities (86-27-05). Improved In-Plant Iodine Instrumentation Under Accident Conditions - Item III.D.3.3

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8.1 Position NUREG-0737, Item III.D.3.3 requires that each licensee shall provide equipment and associated training and procedures for accurately determining the airborne iodine concentration in areas within the facility where plant personnel may be present during an acciden Documents Reviewed The implementation, adequacy and status of the licensee's in plant i' iodine monitoring under accident conditions were reviewed against the criteria in Section 3.0 and in regard to the documents stated in Attachment 1.D. The licensee's performance relative to these cri-teria was determined by:

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Interviews with the radiological protection supervisor, chem-

istry supervisor, members of the Emergency Monitoring Team (Technicians) and training instructors;

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Review of applicable procedures for In-Plant Survey Team emer-gency operations;

Review of applicable lesson plans and training records;

Direct observation of performance during walk-through; and

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Verification of equipment availability and storag .2 System Description i The licensee had assembled adequate numbers of dedicated equipment and supplies, including air samplers, silver activated silica gel

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cartridges, and portable radiation detectors (HP-210 probe connected to E-140 count rate meter) to sample in plant for radiofodine Training on the use of the licensee's procedures appeared adequate.

The licensee developed procedures to be used during an emergency for

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sampling in plant for radiciodines. The procedure provided a calcu-

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lated method to convert filter count activity to an estimated I-131 dose equivalent. Through information provided to the inspector during the inspection, and a subsequent call from the licensee on September 22, 1986, the _ inspector was able to verify the I-131 dose equivalent conversion factors used in the procedur .3 Findings

The licensee meets the requirements specified in NUREG-0737, Item j III D. . Exit Interview The inspectors met with the licensee management representatives (denoted in Section 1.1) at the conclusion of this inspection on August 22, 1986, to discuss the scope and findings of the inspection as detailed in this

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report. During this meeting, the licensee agreed to provide the schedule

for correcting three inspector findings. These items are identified as 86-27-01, 86-27-04 and 86-27-06 in this report.

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Attachment Documentation For NUREG-0737, II. Millstone Nuclear Power Station Unit 3 FSAR, Sections 9.3.2.6, Post Accident Sampling System; 12.3.1.3, Accident Shielding, and Section 5, Reactor Coolant Sampling

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Millstone Technical Specifications, Section 6.8, Procedures and Programs

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Safety Evaluation Report, MIREG-1031, July 1984, Section 9.3. Station Procedures EP-4224, Revision 2, " Reactor Coolant and Liquid Waste Post Accident Sampling" EPIP-4225, Revision 1, " Ventilation and Containment Air Post Accident Sampling" EPIP-4226, Revision 0, Unit 3 Core Damage Estimate Procedure" CP 3800C, Revision 0, " Containment Air Post Accident Sampling System Training and Sample Acquisition Procedure" CP 3800A, Revision 0 " Reactor Coolant PASS Training and Sample Acquisition Procedure" CP 3801A, Revision 0, "P-E ICP/AA/HGA Spectrophotometer" CP 38018, Revision 1, "P-E Sigma 2000 Gas Chromatograph" CP 38010, " Computer Radioisotopic Analysis System" CP 801/2801Y, Revision 1, " Ion Chromatography" SP 3885, Revision 0, " PASS Reactor Coolant Operability Test" SP 3886, Revision 0, " PASS, Containment Air Operability Test"

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Phase 1 Testing Procedures T3311CIE01, " PASS Panels 3 SSP-PNL1 and SAS1" Test Results Approved October 14, 1985 T3311CIE02, " PASS Panels 3 SSP-PNL2 and SAS2" Test Results Approved October 15, 1985 T3311C1E03, " PASS Flush Pump 3 SSP-P3" Test Results Approved November 7,1985

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T3311CIE04, Containment Air' Sample Compressor" Test Results Approved November 6, 1985.

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License Memoranda t

Letter.from General Dynamics to Northeast Utilities Service Company, dated

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May 24, 1982, re: proposal to provide PASS Spare Parts Package Letter from J.' F. Opeka (NNECO) to V. Noonan (NRC), dated January 24, 198 Internal Correspondence M. Tortora to J. Waters, dated July 12, 1985, re: PASS Sample Points'

M. Tortora to J. Waters, dated August 14, 1986, re: Isotopic

Representation in Diluted Sample M. Tortora to J. Waters, dated August 21, 1986, re: PASS Audit of Time Constrain .

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Drawing S&W Dwg. No. 12179-EM-155A-5, dated April 24, 1986

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Attachment Documentation for NUREG-0737, II.F.1-1&2 NRC Documents and Correspondence

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M. Shanbaky, Chief FRPS, Region 1, to R. Bellamy, Chief EPRPB, Region 1, dated May 22, 1986

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R. Baer, Chief EGCB to R. Bellamy, Chief EPRPB, Region 1, dated April 15, 1986

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T. Martin, Dir. DRSS, Region 1, to J. Opeka, Senior Vice President, dated October 29, 1985 Northeast Nuclear Energy Corporation Drawings

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Piping and Instrumentation Diagram, Reactor Plant Ventilation, Dwg. No. 12179-EM-1480-4

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Piping and Instrumentation Diagram, Reactor Plant Ventilation, Dwg. No. 12179-EM-148A-3B

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Piping and Instrumentation Diagram, Reactor Plant Ventilation, Dwg. No. 12179-EM-148E-1A

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Wiring Diagram Radiation Mon. Sys. SH.32 Dwg. No. 12179-EE-11L-6

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480V MCC One Line Diagram Rod Control Area S Dwg. No. 12179-EE-1AQ-17 Kaman Instrument Corporation - Correspondence and Documents

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A. Colton, Pres. to Dir. I&F, dated February 27, 1986

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F. Bergamo, Cust. Ser. Mgr., to owners of KMG-HRH, dated March 10, 1986

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Technical Newsletter, from T. Johnson, Sept./Oct. 1983, Vol. 1, No. 1

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Operation / Maintenance Instructions and Parts Catalog for Millstone Nuclear Power Station - Unit 3 Radiation Monitoring System (MG24),

April 6, 1983 Northeast Nuclear Energy Corporation Emergency Plan Implementing Procedures

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EPIP 4225, Rev.1, " Unit 3 Ventilation and Containment Air Post Accident Sampling"

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Attachment m

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EPIP 4225, Rev. O, " Unit 3 Ventilation and Containment Air Post Accident Sampling"

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EPIP 4701, Rev. 2, " Unit Incident Assessment, Classification and

Reportability"

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EPIP 4701-3, Rev. 7, " Emergency Action Levels"

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EPIP 4201, Rev. O, " Radiological Dose Assessment" I Northeast Nuclear Energy Corporation Calibration Procedures

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Millstone Unit 3 Ventilation Stack (Turbine Building) Channel Calibration, SP3449A01, Rev. 1, November 16, 1985

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Millstone Unit 1 Mainstack Channel Calibration, SP3449801, November 16, 1985

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Main Steam Header Radiation Monitor 3 MSS *RE75, 76, 77, 78, 3490B12, Rev. O, May 5, 1986 I -

Auxiliary Feedwater Pump 3FWA-T1 Exhaust Radiation Monitor, 3490812,

Rev. O, May 5, 1986 j

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Ventilation Vent Flow Rate, SP3449A21, Rev. O, November 9,1985

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l Licensee Documents

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Millstone Unit 3 Radiation Monitor Manual, Alert / Alarm Set Points and Conversion Factors, Main Steam Line Relief Line Monitoring

, 3MSSRE 75, 76, 77, 78 and Turbine Driven Auxiliary Feed Water Pump j Steam Exhaust 3MMRE79, January ~ 15, 1986

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Plant Incident Report, "High-Range Rad. Monitor Saturation Problem",

dated April 14, 1986

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J. Crockett, Supt., to R. Laudenat, Lic., dated May 12, 1986

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J. Crockett, Supt., to F. Bergomo, Kaman Inst. Corp. dated May 12, 1986 i

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Millstone Nuclear Power Station Unit 3 Final Safety Analysis Report

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Millstone Nuclear Power Station Unit 3 Technical Specifications Northeast Nuclear Energy Abnormal Operating Procedures i -

A0P 3573, Rev. 7, " Radiation Monitor Alarm Response"

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A0P 3556, Rev. 7, " Steam Generator Tube Leak"

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Attachment Documentation For NUREG-0737, II.F.1-3 Millstone Nuclear Power Station Unit 3 FSAR, Section 12.3.4, Area Radiation and Airborne Radioactivity Monitoring Millstone Unit 3 Technical Specifications, Sections 3/4.3.3.6, Accident Monitoring Instrumentation

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Station Procedure SP3449H21, Rev. O, " General Atomic Containment Area

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High-Range Channel Calibration" Station Procedure SP3449H01, Rev. O, " Containment High Range Channel i Calibration" Station Procedure SP407D, Rev. 6, "Drywell High Range Radiation Monitor

Calibration"

, Station Procedure SP407E, Rev. 2, "Drywell High Range Radiation Monitor / Detector Calibration"

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j Station Procedure SP2404AY, Rev. O, " Containment High-Range Monitors RIT 8240, 8241 Calibration" Station Procedure EPIP 4212, Rev. 4, "Drywell/ Containment Curie Level Estimation" ,;

General Atomic Report No. E-255-978, " Energy Response Test and Dose Rate Calibration of Model RD-23 High-Range Radiation Monitor Detector" Kaman Instrumentation Corp. " Report of Calibration, Model KDA-HR High-Range s

, Area Monitor" .

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Stone & Webster E&DCR F-E-14, 862, page 7 and 8 of 12, Containment Structure Stone & Webster E&DCR N-EC-01786, page 5 and 6 of 26, Containment Building El 51' 4" Stone & Webster E&DCR N-EC-01778, page 6 and 7 of 16, Containment Structure Licensee Event Report No. 50-423/86-023-00 Licensee Special Report dated January 24, 1986, " Containment Area High-Range Radiation Monitors"

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Attachment Docume ation for NUREG-0737, III.D. i Millstone Nuclear Power Station Unit 3 FSAR, Section 12.5

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Station Procedure EPIP 4203, "EMT #4 - In-Plant Radiological Sampling and

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Monitoring, Revision ;

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NUREG/CR-0314, "An Air. Sampling System for Evaluating the Thyroid Dose

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Commitment due to Fission Products Released from Reactor Containment"

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Attachment II Comparison of Analytical Results A. Chemical Analysis Boron This procedure calls for an initial 10' dilution of reactor coolant samples from the PASS system, which was replicated for the analyses of the test samples. The test data were:

Standard Dilution Factor Concentration Found In ppm i 60 10,000 5032 ppm Acceptance Criteria > 1000 ppm i 5%

Results were in agreemen Chloride This procedure calls for a 25 ml dilution prior to analysi The test data were:

Standard Analysis Acceptance Criteria 2 .1 ppm 2 %

37.4 1 1.2 ppm 3 %

8 .2 ppm 7 i 10%

Results were in agreemen gH In-line measurement compared to large volume measurement:

As Read Analyzed In Lab Acceptance Criteria .48 t 0.3 pH units Results were in agreemen ,

B. Isotopic Analysis The identifiable nuclides in the reactor coolant and containment atmosphere samples were compared to the normal sampling methods. The acceptance criteria was within a factor of The results were in agreemen _. ._