IR 05000293/1987035
| ML20237A581 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 12/07/1987 |
| From: | Bicehouse H, Davidson B, Kirkwood A, Kottan J, Pasciak W, Struckmeyer R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20237A567 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 50-293-87-35, NUDOCS 8712150108 | |
| Download: ML20237A581 (38) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-293/87-35 Docket No.
50-293 License No.
DPR-63 Priority - Category C Licensee: Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 Facility Name: Pilgrim Nuclear Power Station Inspection At: Plymouth, Massachusetts Inspection Conducted: October 5-7, 1987 l Inspectors: d d _ /)/r/87 J. J. Kottdn', Laboratory Specialist datt / cA A AR m/ol/v7 l A. S. KirJwoot, Radiation Specialist date' ) ! L lA f, ) ~ R. K. Struckmeyer,fhdiation Specialist date l7~ .? ) l ' ,, _.
, l B.u S.d'Dsvidson, Radiation Specialist date / / ' I ? ',7!9 9 . . m H. J. Bicehous e adiation Specialist date Approved by: _ . e.
u rf_ (2 9 @% W. J.j P 'scialG ' Chief, Effluents Radiation date Pro't tion Section l Inspection Summary: I l Areas Inspected: Special, announced safety inspection of the following: I licensee action on previous findings; radiological controls organization and l staffing; radiological measurements laboratory; nonradiological measurements laboratory, whole-body counting laboratory; actions on previous NUREG 0737 ?DR#A&K05000293 B B71300 l G ' PDR _ - _ _ _ _ _ _ _ _ _ _ _ _ _
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g findings; transportation activities; 10 CFR 61 compliance; radiological efflu-ents; gaseous and liquid radioactive waste processing systems; and radiological environtrental sampling.
Results: Two violations were identified (failure to properly identify and quantify radionuclides contained in a radioactive waste shipment; details Paragraph 5.4 and 5.5.
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i . DETAILS ! 1.
Persons Contacted i 1.1 Licensee Personnel '
- C. Gannon, Chief Radiological Engineer
- B. P. Lunn, Senior Compliance Engineer
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- S. L. Bibo, Audit Group Leader
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- J. C. Mattia, Surveillance Group Leader i
- S. D. Hudson, Operations Section Manager
- L. E. Whittenberger, R05 Group Leader
- P. Hamilton, Compliance Group Leader
- K. Roberts, Nuclear Operations Manager
- R. Canalas, Chief Chemical Engineer
- R. P. Jens, Radiological Section Manager A. Shatas, Senior Chemical Engineer C. Grevenitz, Senior Chemical Engineer C. Bowman, Rad. Engineering Group Leader B. Dionne, Senior Radiological Engineer
.. C. Morrill, Grade-3 Engineer L. Simons, Senior Quality Assurance Engineer D. Stauder, Technician, General Test Division l C. Goddard, Chemical Engineer l A. Nogas, Chemistry Technician ! l R. Bowen, Chemistry Technician l l D. Fountain, Chemistry Technician ! D. Mitchell, Chemistry Technician L. Loomis, Chemistry Technician J. McCarthy, Chemistry Technician, Training J. Spangler, Nuclear Training Specialist ' J. Hurley, Senior Health Physics Technician a R. Green Quality Assessor W. Deacon, Assist. to Sr. VP, Nuclear , ! J. Henderson, R. P. Technician Other licensee personnel were contacted or interviewed.
1.2 Contractor Personnel D. Alt, Senior Health Physics Technician 1.3 NRC Personnel
- J. Lyash, Resident Inspector, Pilgrim
- C, Oberg, Radiation Specialist, Region I
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- J. Lee, Radiological Engineer, NRR l
C. Warren, Senior Resident Inspector, Pilgrim T. K. Kim, Resident Inspector, Pilgrim
- Attended the exit interview on October 9, 198 _ - _ _ _ _ _ _ _ _ _ _ _
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2.
Scope of the Inspection This special, announced inspection reviewed the licensee's solid radioactive (radwaste) preparation, packaging,andshipmentprogram; radiological, nonradiological and whole-body counting laboratories; actions taken on previous NUREG 0737 findings; the licensee's program for 10 CFR 61 compliance, radiological effluents (gaseous and liquid waste processing systems), radioanalytical environmental sampling, and radiological controls organization and staffing as it pertains to these areas.
3.
Previously Identified Items (0 pen) Follow-up Item (50-293/85-27-02): Demonstrate the adequacy of the
sample dilution method and related equipment. The dilution method cannot j be demonstrated until after plant start-up.
The licensee stated that zinc analyses would probably be used to arrive at the dilution factor.
Based on the results of the zinc standards analysis performed during this inspection, this appears to be a viable approach. This will be reviewed during a subsequent inspection.
(Closed) Follow-up Item (50-293/85-27-09): Boron analysis by carminic acid method. The licensee obtained results which were within plus or minus ten percent when using the 1-8 ppm calibration curve.
(See Section 11.) The licensee has implemented a procedure change notice to Procedure 7.11.1, " Analysis of Liquid Samples for Boron (by spectrophotometry) Under Accident Conditions", which incorporates the 1-8 ppm calibration.
(0 pen) Follow-up Item (50-293/85-27-10): Clarify commitments and capa-bilities relative to pH analysis.
The inspector reviewed licensee documentation, and noted that the licensee had not yet formalized his position on this matter.
This will be reviewed during a subsequent inspection.
(0 pen) Follow-up Item (50-293/85-27-27): Licensee to clearly establish high range noble gas monitor readout range and monitor overlap.
The licensee stated that some work had been done on this item, but it had not been completely reviewed and resolved.
This will be reviewed during a subsequent inspection.
(0 pen) Follow-up Item (50-293/85-27-30): Post-accident sampling of the nain stack and reactor building vent. An Engineering Service Request (ESR) has been initiated to address this item.
The ESR has not yet been completed.
This will be reviewed during a subsequent inspection.
(0 pen) Follow-up Item (50-293/85-27-31): Personnel exposure during sam-pling of effluent release paths.
The personnel exposure studies cannot be completed until Item 50-293/85-27-30 above is completed.
This will be reviewed during a subsequent inspection.
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- (0 pen) Follow-up Item (50-293/85-27-33): Capability of the reactor building vent to collect representative samples. Action on this item has been initiated by the licensee, but it is not yet complete.
This will be reviewed during a subsequent inspection.
(0 pen) Follow-up Item (50-293/85-27-34): Capabilities of the sampling system to collect representative samples. Action on this item has been . initiated by the licensee, but it is not yet complete.
This will be reviewed during a subsequent inspection.
' (0 pen) Follow-up Item (50-293/85-27-35): This item consists of two sub-items: shielding for effluent sample holders and purging of effluent l sample canisters prior to analysis.
The licensee has begun action on both ' of these items, but it is not yet completed.
This will be reviewed during a subsequent inspection.
' (Closed) Inspector Follow-up Item (219/86-09-02): Determine distance and direction for each sample location and make corrections to technical specifications and ODCM as necessary.
The licensee checked the loca-
tions of all of its environmental monitoring stations and revised the
, ! descriptions of these locations as necessary in its ODCM and in relevant ! procedures.
The radiological engineering group had prepared a change j request to amend the description of sampling locations in the technical specifications, but the licensee now would prefer to eliminate the actual locations from the technical specifications, and refer to the ODCM instead.
The inspector noted that this change could not be implemented immediately; in the meantime, the functioning of the environmental monitoring program should not be impaired by strict adherence to the locations stated in the , technical specifications while the change is being sought.
l (Closed) Unresolved (50-293/86-35-01): Review purchase requisition docu-mentation for specifications on charcoal absorbents used in the Standby l Gas Treatment System (SGTS) and to ensure Quality Assurance criteria are ' provided.
The inspector reviewed the licensee's specification for the , charcoal used in the SGTS ventilation system and found it adequate.
This l item is closed.
' ' l (Closed) Violation (50-293/86-10-01): Failure to provide Quality Control
Program to ensure compliance with 10 CFR 61.56.
The licensee's actions (as described in the licensee's letter dated June 20,1986) were reviewed.
The licensee revised a Quality Control Instruction to require annual sur-veillance of dewatering processes and implemented the surveillance at more frequent intervals.
In addition, the licensee's on-site Quality Control Group provided inspections of container loading, dewatering and shipping preparation activities.
This item is closed.
(Closed) Follow-up Item (50-293/86-10-02): Procurement of transportation services expired.
The licensee had identified deficiencies regarding procurement of transportation-related items and services without a current purchase order during a Quality Assurance audit.
The licensee reviewed active purchase orders to ensure that quality-related items and services l L--___-.---
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were included and established an on-site Procurement Quality Assurance function to oversee the area. This item is closed.
(Closed) Follow-up Item (50-293/86-10-03): Resolve labeling conflict with' shipping container supplier. The licensee corresponded with the supplier i to clarify markings on the Chem-Nuclear Systems Inc. shipping container.
The supplier replied that changes to the appropriate Certificates of Comp'iance had been requested to include the specific model numbers used l by the supplier. This item is closed.
(Closed) Violation (50-293/86-10-04) Failure to follow procedures to record surveys and total activities for shipments. Actions described in the licensee's letter dated June 20, 1986 were reviewed. The incorrect total activity on Shipment.No. 85-79 was corrected, technicians received' . additional instructions regarding surveys and technician. assignments to shipping activities were increased as described in the licensee's i response. This item is closed.
4.0 Organization and Management Controls i 4.1 Radiation Protection Organization g ruments Reviewed: Technical Specification 6.0 - Health Physics Staff Position Descriptions - Regulatory Guide 8.8, "Information Relevant to Ensuring that - Occupational Radiation Exposures at. Nuclear Power Stations Will be as Low as is Reasonably Achievable" . 4.1.1 Description The formally established radiation protection organization for the Pilgrim Nuclear Power Station is depicted in Figure'1.
This organization chart reflects a clearly defined reporting chain which appears to be functional. The position identified in Figure 1 as " Radiological Section Manager" is responsible for the overall planning, development and implementation of all radiological health programs at the station.
The position description states that the incumbent fulfills requirements ~ for Radiation Protection Manager (RPM) as specified in Regulatory Guide 1.8, however, the Chief Radiological Engi-neer currently holds the RPM title.
This does not affect the reporting chain and appears to be for administrative purposes only. The Chemistry Lab is under the Operations Section Manager.
It appears to-be organized under a defined chain of command.
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Based on the above findings, this portion of the licensee's prcgram is adequate.
4.1.2 Scope of Responsibilities and Authorities.
I Figure 2 depicts the functional positions within the radiation protection organization (Radiological Section).
As shown in Figure 2, the Radiological Section is made up j of four distinct sub-groups [ Health Physics /Radwaste under i the Chief Radiological Engineer, On-Site (ALARA) and i Off-Site Technical Support as well as the Operations Support Group (Instruments / Dosimetry)]. The following position descriptions were reviewed to determine if the scope of responsibility and lines of authority were clearly defined: l l Radiological Section Manager - The areas of assigned responsibility and reporting appeared to be clearly _ defined and adequate, with the aforementioned RPM exception.
i - Chief Radiological Engineer - Has responsibility for , the implementation and coordination of the Pilgrim Station Health Physics Program.
Responsibilities and reporting chain appear to be appropriate, with the j l exception that the Operations Support Group
I maintains, calibrates and supplies radiation surveillance equipment.
I - Senior Supervising Radiological Engineer - Plans and
coordinates work in the areas of radiological l surveillance; radiation work permit / radiation protection job coverage; dosimetry / respiratory ' protection assessment needs and before and after , l decontamination surveys.
Contrary to the position description provided (dated 16 Oct 86), the incumbent I no longer is responsible for calibration of radiation surveillance equipment and the issuance, testing and , repair of respiratory protection equipment.
- Suaervisor - Radiological - Four supervisors are respon-si)le for assuring the tinely completion of various health physics duties broken down into different plant areas and general RWPs.
These are first level supervisory positions.
__ Radiation Protection Technician - Directly performs'and documents the results of radiation surveys.
Contrary to the 16.0ct 86 position description, the incumbent no longer calibrates station radiation monitors or performs decontamination work (vendors used).
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) l l Minimum qualifications listed in such areas as Mechanics, Chemistry, Electronics and Nuclear Engineering may be unrealistic.
Based on the above review, it is apparent that several func-tional positions within the Radiological Section have areas of l responsibility no longer valid or listed qualification criteria not entirely accurate.
Excluding the above, the Radiological Section Organization was found to be satisfactory.
However, it is recommended that position descriptions be developed and . written in a manner to clearly define the job position ) areas of responsibility.
Additionally, a means should be I ' provided to periodically update the job descriptions, Based on observations made during the appraisal period and interviews with plant personnel, it appeared that the
! Radiological Section had adequate station management support.
4.1.3 Staffing Under the Radiological Section Manager, the Chief Radio-logical. Engineer (HP Group Leader) has a staff of 36 ' people for the day-to-day implementation of the overall < l radiological control program. The health physics staff consists of (1) Senior Engineer; (4) Supervisors to
implement the in-plant health physics program; (26) health ) physics technicians; and (1) health physics clerk.
Also, j contractor supplied health physics technicians are used to d stpplement the health physics group on an as-needed basis.
! The number of technicians has doubled since the last I l Health Physics Appraisal (1980).
l l Radwaste functions are implemented by the Senior Engineer and Supervisors, and are performed by contract / vendor personnel.
Health physics coverage at the Pilgrim Station is on a 24-hour basis. There are two health physics technicians assigned for each weekend 12-hour shift. All of the technicians rotate on an assigned basis through the day, night and weekend shifts.
Rotation of permanent ' technicians is restricted by a management document dated January 12, 1987 to periods of not less than 2_ months.
Health physics technicians report to one of four supervisors who assigns them specific areas of responsibility for the day-to-day implementation of the health physics program.
l During nonroutine periods, such as a refueling outage, as ' many as 120 additional staff may be needed and are normally . obtained through contractors.
Typically, 60 to 65 are !
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requested and 40 to 45 are obtained.
A plan to increase the number of permanent technicians to 56 has been submitted to management to overcome contracting problems.
4.1.4 Audit Program Documents Reviewed - Boston Edison, QA Manual, Volume II - PNPS Technical Specification 6.0 - QAP Procedure No. 2.04, " Annual Independent Review of BEco's QA Program , , ' - NOD Procedure 1.3.26, " Response to Deficiency Reports..." - Audit Reports l 86 Chemistry 87 Chemistry Hydro Nuclear Audit for the BECo Radiological Improvement Plan (RIP), February 28, 1985 Hydro Nuclear Audit for the BECo Radiological - Improvement Plan, September 1986 Radiological Section Task Assignment Sheets - 86-136 ' 86-138 86-139 - Office Memorandum, Subject: "QAD Audits of BEco Nuclear Organization," dated November 21, 1986 BECo has established a formal internal audit function for radiological matters as a part of their Quality Assurance (QA) Program. Audits are performed by the Corporate QA staff (see Fig. 3) with technical support by specialists from other BEco Departments or outside consultants.
The following related audit subject areas and frequencies were identified i in Technical Specification 6.5 and Office Memorandum dated November 21, 1986:
. ' ! Subject Frequency i Radiological Environmental Monitoring Annual ' Chemistry Annual Health Physics Annual Radioactive Waste Shipments Annual Copies of these audit reports were distributed to the various section managers, as well as the Vice President - Nuclear, who received a copy of all audit reports.
Audits were conducted in accordance with approved audit schedules and audit plans.
The audit plans were developed
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from review of the previous audit, technical specifica-tions, governing procedures, regulatory guidance and requirements and QA program criteria.
Review of several recent audit plans and reports indicated that the audits were generally thorough and well-documented.
Items identified which require responses (deficiency reports).
were entered into a tracking and responsibility assignment system and response follow-ups were conducted.
Findings related to safety or meeting certain other criteria demanding quicker response are further designated as "si gni fi ca nt. " Recent audit reports and associated Deficiency Reports were reviewed to determine responsiveness of audited , operations to audit findings.
In general, responses to audit findings appeared to be timely.
The method to request an extension of a required response, Vice President - Nuclear approval, did not appear to be over used.
Two comprehensive radiation protection audits done by an outside consulting firm were conducted in February 1985 and September 1986.
Three areas were looked at and tracked for , signs of significant improvement.
A lack of formal defini-l tion within position descriptions was noted in j ' February 1985.
By September 1986, position descriptions j had been written defining responsibilities and reporting j chains but first-line supervisors had not reviewed their { own. This was corrected by a Radiological Section Task j Assignment dated October 20, 1986.
In the earlier audit, i a need to clarify the assignment of responsibility in the ) radioactive waste management area was corrected by the i time of the following audit.
In 1985 a need for a ) management audit program was identified as well as a need for a means to track problems.
By 1986, this was rede-fined to mean an internal audit function.
Currently, safety problems are tracked on a computer printout titled Radiological Section Task Assignment (RSTA).
An interim measure to audit the radiation protection program is being conducted by the Independent Radiation Protection
Oversight Committee (IROC).
In coordination with an onsite assessor who reports both to the committee and the Senior Vice President - Nuclear, the IROC meets on site ' monthly to assess the effective implementation of their recommendations for an upgraded Pilgrim Nuclear Power , Station Radiation Protection Program.
The committee is composed of four natior, ally recognized experts in the health physics and nuclear power fields.
It is planned that the internal audit function will be assumed by the Radiological Section early in 1988 with the addition of new staffing.
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' BECo's QA internal audit program appeared to be highly structured, both corporate and.onsite, and carried out in a thorough, competent manner.
In the limited sample of responses to audit findings reviewed, some requested delays were observed but it was considered to be a.small , portion and not indicative of lack of full management j support of the audit program. The use of outside j qualified technical specialists to supplement the QA { auditor in assessing the adequacy of an activity is considered to be a strong point in the licensee's internal audit program.
This program should serve as a good management control tool for evaluating and upgrading the radiation protection program provided adequate delineation l I of safety-related items is made and a high priority given to resolving these deficiencies as quickly as possible, without unnecessary extensions.
i 5.0 Solid Radwaste Transportation and n_ CFR 61 The inspector reviewed the license el solid radioactive waste (radwaste) processing, preparation, packaging sd shipping program (as implemented by the licensee) from April 1,1986 through October 8,1987. A sample . of twelve radwaste shipments was selected (representing Low Specific l Activity and Type A quantities under Title 49 and classes A, B and C l under 10 CFR 61) and reviewed relative to requirements in 10 CFR 20.311, 10 CFR 61.55-56, 10 CFR 71, 49 CFR 170-189 and the licensee's Technical Specifications.
Each shipment was reviewed and discussed with the licensee to assess the overall adequacy and effectiveness of the licen-see's program.
Extensive changes to the program had been made by the licensee in early 1987 to increase quality assurance / quality control (QA/QC) activities in the program areas reviewed.
The effectiveness of these changes was also assessed.
5.1 Managemant Controls The organizational structure of the licensee's program for proc-essing, packaging and shipping radioactive materials including responsibilities and interfaces of the Quality Assurance, Nuclear ! Engineering and Nuclear Operations Departments was reviewed. The licensee extensively changed and. improved the QA/QC aspcts of the program in a complete revision to Nuclear Organization Procedure (N0P) 87RC1 (dated September 26,1987).
NOP 87RC1 clearly defined the roles and responsibilities of the departments.
The licensee's procedures for processing, preparation, packaging and shipping solid radwaste materials were reviewed relative to the criteria above.
The following procedures were reviewed, discussed with the licensee and checked for implementation: l -__ _ _ _ _ _ _ _ _.
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- Procedure No. 6.9-160, " Shipment of Radioactive Material (RAM)," . ! Revision 23 (August 7, 1987); - - Procedure No. 6.9-186, "High Integrity Containers," Revision 3 (Septemoer 24,1987);
- Procedure No. 6.9-193, "10 CFR 61 Classification of Radwaste," l Revision 3 (January 21,1987); Procedure No. 6.9-200, "00T Classification of Radioactive - Material," Revision 5 (March 9, 1987); - Procedure No. 6.9-211, "10 CFR 61 Sampling, Revision 1 (September 16,1987); Procedure No. 6.9-212, " Handling and Loading Type 'A'. Shipping Casks," Revision 0 (September 23,1987); _ Procedure No. 6.9-213, " Handling (and Loading Procedure for Type 'B' Shipping Casks," Revision 0 September 23,1987);and - Temporary Procedure No. TP 86-171, "Desludging, Decontamination and Solidification of the Dryer Separator Pool Contaminates," - Revision 1 (February 27,1987).
The procedures were adequate and were being implemented by the licensee.
The licensee's training program was briefly reviewed relative to i commitments in the licensee's response to NRC-IE Bulletin No. 79-19.
Selected individuals in the Radiological Controls, Quality Assurance and Qualit" Control organizations were interviewed regarding their ongoing training.
No deviation from commitments were noted in this review.
5.2 Quality Assurance / Quality Control The application ot~ the licensee's QA/QC program to the licensee's responsibilities as a radwaste generator and shipper was reviewed.
Procedures were reviewed to determine if appropriate quality control
becks were included to ensure proper classification of radwaste shipments under 10 CFR 61.55, proper radwaste form under 10 CFR l 61.56,* .fication of appropriate attributes in receipt inspections of shipping containers and disposal packages and verification of dpplicable items of licensed packaging's Certificates of Compliance.
Each shipment was reviewed relative to procedures (in force at the time of shipment) to determine that the appropriate quality control checks had been implemented. QA audits and surveillance were reviewed and discussed with the licensee to determine the attributes of the program reviewed, management review and resolution of findings and recommendations and the qualifications of the audit and surveil-lance personnel.
Within the scope of this review, no violations were noted. The licensee had substantially upgraded quality control activities in 1987 addressing weaknesses and problems noted in Inspection No.
50-293/86-10 and this inspection (see related items below).
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5.3 Radiochemistry /Part 61 Sampling The inspector briefly reviewed licensee fuel performance and general radiochemical data related to fission and activation product behavior to determine the general adequacy of the licensee's sampling and vendor analysis program for difficult-to-identify radionuclides.
l Although fuel performance data indicated no fuel pin leakage since 1980, the persistent presence of Cesium-137 and Strontium-90 in the waste stream analyses suggested a source of fission product intrusion.
The licensee was unable to explain the apparent discrepancy in the data. As noted during Inspection No. 50-293/86-10, the licensee had sampled resin shipments individually and used actual analytical data from the vendor to assess hard-to-identify radionuclides.
5.4 Radwaste Generator Requirements The twelve shipments were reviewed for compliance with the following i radwaste generator requirements: - Waste Manifests under 10 CFR 20.311(d)(4) and 20.311(b) and (c); _ Waste Classification under 10 CFR 20.311(d)(1) and 10 CFR 61.55; - Waste Form and Characterization under 10 CFR 20.311(d)(1) and 10 CFR 61.56; _ Waste Shipment labeling under 10 CFR 20.311(d)(2) and 10 CFR 61.55; ~ Trackingofwasteshipmentsunder10CFR20.311(d),(e),(f) ) and (h); and { - Adherence to disposal site license conditions under 10 CFR 30.41.
Within the scope of this review, the following apparent violation j was noted: 10 CFR 20.311(b) requires, in part, that each shipment be ~~ accompanied by a waste manifest that describes as completely as practicable the total radioactivity in the shipment.
10 CFR 20.311(c) requires a certification by the waste generator that the shipment is properly described.
Contrary to these requirements, the licensee sent a shipment of noncompactable trash to Barnwell, South Carolina, on or about June 27, 1986 as licensee's shipment No. E6-44.
The waste mani-fest accompanying the shipment listed a total radioactivity of approximately 0.1 millicurie when the actual total radioactivity was approximately 120.8 millicuries.
The licensee's authorized representative certified that the shipment was properly described when its total radioactivity was in error. The error was dis-covered by the site operator and subsequently corrected on the i
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i i waste manifest.
Listing an incorrect total radioactivity on j the waste manifest for shipment No. 86-44 and certifying that j the shipment was properly described are an aggregate violation ) of 10 CFR 20.311(b) and (c).
50-293/87-35-01.
5.5 Radwaste Shipping Requirements The twelve shipments were reviewed relative to criteria contained in 10 CFR 71 and 49 CFR 170-189 to determine if transportation j requirements associated with procurement and selection of packages, i preparation of packages for shipment and delivery of the packages to carriers had t,een met.
Within the scope of this review, the following apparent violation was noted: ~~ 10 CFR 71.5(a)(1)(vi) requires preparation of shipping papers in accordance with 49 CFR 172, Subpart C.
49CFR172.203(d)(i) requires the name of each radionuclides in the shipment, 49 CFR 172.203(d)(iii) requires the activity of each package and 49 CFR 172.204(a)(1) requires a statement that the shipment is properly described in the shipping papers.
Contrary to these requirements, the' licensee sent a shipment of dewatered powered resin from the licensee's Reactor Water Cleanup -(RWCU) System to Barnwell, South Carolina, on or about May 1, 1986 as licensee's shipment No. 86-34.
Vendor i l analyses of resins showed that Iron-55 was present in RWCU resins.
However, the licensee failed to include the name of
that radionuclides, its activity in the package and certified that the shipment was properly described.
Shipment No. 86-34 contained approximately 296 curies of radionuclides other than Iron-55.
Calculations made by the inspector indicated
an additional 40-100 curies of Iron-55 were also present but unlisted.
Failure to name Iron-55, include its activity in the shipment and stating incorrectly that the shipment was i properly described constitute an apparent violation of 10 CFR 71.5(a)(1)(vi).
50-293/87-35-02.
l 5.6 Transportation Incidents The licensee stated that no shipping problems had been noted and no violations or warnings had been received from the State Regulatory agencies regarding their shipments during the period reviewed.
6.0 Chemistry Laboratories This part of the inspection assessed the capability of the licensee to perform chemical and radiochemical measurements.
This assessment con-sisted of a review of the licensee's organization, training, facilities and equipment, as well as the submission of chemical and radiochemical standards.to the licensee for analysis.
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l 6.1 Organization The licensee's chemistry organization is a part of the station operations department. The Chief Chemical Engineer reports to the , Operations Section Manager. Under the Chief Chemical Engineer are two Senior Chemical Engineers, one Chemical Engineer, one Chemistry Supervisor, Chemistry Technicians, Technicians in Training, and Apprentice Technicians.
6.2 Training Chemistry Technician training consists of an initial chemistry technician training program and a chemistry technician apprentice program.
The initial chemistry technician training program is reserved for those technicians who meet ANSI N18.1-1971, " Selection and Training of Nuclear Power Plant Personnel," experience requirements when hired.
This training is 14 weeks in length and includes both theory and procedure qualification. The chemistry technician apprentice program which is two years in length is intended to train and qualify personnel who do not meet the ANSI criteria when hired.
In addition, a continuing training program is maintained for quali-fied technicians.
This retraining activity is scheduled for one week out of every ten weeks.
No violations were identified in this area.
6.3 Facilities and Equipment The licensee's chemistry laboratory consists of two rooms, containing fume hoods and laboratory benches, which are used for chemical and radiochemical analyses. The counting room is located in a separate room off the laboratory. The licensee possesses state-of-the-art analytical instrumentation including: a computer based gamma spect-roscopy system and germanium detectors, an ion chromatograph, and an AA/ICP spectrometer unit. Other laboratory supplies such as pH meters, analytical balances, glassware, reagents and chemicals appeared to be in adequate supply.
In addition, the licensee possesses a computer based laboratory data management system used for maintaining and trending laboratory generated data.
6.4 Radiological Standards Results ' During the inspection, radioactivity standards were submitted to the licensee by the inspector for analysis. The standards are used to evaluate the licensee's capability to measure radioactivity in both i effluent and in-plant samples as required by Technical Specifications.
The standards duplicated the types of samples and nuclides that the licensee would encounter during normal operation.
The licensee used routine methods and equipment for analysis of the standards, i _ - - D
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t - The results of the standards measurements comparison indicated that
all of the measurements were in agreement under the criteria used for comparing results.
(See Attachment 1.) The inspector submitted charcoal cartridges containing radioactivity distributed both uniformly and on the face of the cartridge. Although the licensee calibrates with only a face loaded cartridge, the licensee made cali-bration adjustments in order to measure the uniformly loaded charcoal cartridge. This indicated the licensee's ability to measure radio-activity on charcoal cartridges regardless of the distribution of the radioactivity within the cartridge.
During routine operations, however, the licensee normally encountered only face loaded charcoal l cartridges.
In addition, the licensee stated that both face loaded and uniformly loaded cartridges would be used for future calibrations.
The data are presented in Table 2.
6.5 Chemical Standards Results During the inspection, standard chemical solutions were submitted to the licensee for analysis by the inspector.
The standard solutions were prepared by Brookhaven National Laboratory (BNL) for the NRC Region I, and were analyzed by the licensee using normal methods . and equipment. The analysis of standards is used to verify the ' licensee's capebility to monitor chemical parameters in various plant systems with respect to Technical Specification and other regulatory requirements.
In addition, the analysis of standards is used to evaluate the licensee's analytical procedures with respect to accuracy and precision.
t The results of the standards measurements comparison indicated that, with the exception of three values, all results were in agreement under the criteria used for comparing results.
(See Attachment 2.)
The three results in disagreement were a fluoride value at approxi-mately 20 ppb and two zinc values at approximately 300 and 500 ppb.
The fluoride value, although in disagreement under the NRC criteria for comparing results, was within the control limits on the licensee's I fluoride control chart for precision at the 20 ppb level - approxi-mately plus or minus 15 percent. Also, both zinc measurements were with 10% of the NRC known value and the disagreements were not judged I to be significant. These data are presented in Table.3.
In addition to the above standards, the inspector also submitted boron standards to the licensee for analysis using the Post-Accident Sampling System (PASS) procedures.
PASS accuracy requirements for l boron analyses are plus or minus ten percent.
The licensee's pri-mary method of PASS boron analysis is plasma spectrometry. All results were within the required limits using this method. The licensee's backup method is a spectrophotometric method using carminic acid. The initial results using this method were outside the required limits at the one ppm concentration.
These measure-ments were made with a calibration curve over the range _of 1-15 ppm.
Reanalysis of the standards using a calibration curve in the range _ __ -
- . !
{
{ of 1-8 ppm, the normally expected concentration in PASS samples, resulted in all results within the required limits. The data are presented in Table 4.
I i The results of spiked samples which were prepared during a previous ' inspection on February 24-27, 1987 (Inspection Report No. 50-293/ 87-14) are presented in Table 5.
No comparison criteria are used for spiked samples.
I 6.6 Laboratory QA/QC The inspector reviewed the licensee's laboratory QA/QC program as detailed in procedures SI-CH.0100, " Quality Control of Chemistry Laborate"v Data," and SI-CH.0150, " Quality Control of Counting Room Instrume.u ition."
Both procedures address the use of control charts, calibration techniques, standard laboratory practices, and training. In addition, both procedures also address an interlaboratory QC program with specified criteria for comparing results. The licensee also sends spike samples to his vendor laboratory which is used for performing certain radiochemical analyses of effluent samples.
The inspector reviewed selected laboratory QA/QC data and noted that the licensee appeared to be implementing the laboratory QA/QC program as required by the procedures.
No violations were identified in this area.
7.0 Whole Body Counting program This part of the inspection assessed the capability of the licensee to i adequately perform radiological bioassay using a whole body counting system. A whole body counting phantom containing radioactive sources traceable to the National Bureau of Standards (NBS) was submitted to the licensee for analysis.
The phantom duplicated the nuclides arid the organ burdens that the licensee might encounter during normal operation.
The phantom was analyzed using the licensee's routine methods and equipment.
' 7.1 Comparison of Results The licensee uses a whole body counting system which consists of a
shielded chair and an adjustable germanium detector located in front i of the chair.
The germanium detector is coupled to a computer- ) based multichannel analyzer.
Software is supplied by the system vendor. The comparison results for the lung are based on an average of five measurements and the GI tract results are based on an average of two measurements.
Table 1 contains the results of the intercomparison.
Based on these results, no violations were
identified.
7.2 Procedures and Data . The following proceriures for operation and calibration of the whole body counting system were reviewed: _ _ - _ _ _ _ _ - _ _ _ _ - -
] ! SI-RP-4418, " Operation of the N06700 Intrinsic Germanium Whole - Body Counting Chair" SI-RP-5111. " Calibration of the ND6700 Masse Intrinsic - Germanium Whole Body Counting Chair" These procedures include periodic source and background checks, and the maintenance of control charts to trend changes in detector response and background respectively.
Annual calibrations are per- , formed, and quarterly calibration checks, with acceptance criteria, j are performed. Within the scope of this review, no violations were ] identified.
'
8.0 Radioactive Effluents Control Program j l 8.1 Sampling and Radiochemistry The licensee's radiochemistry program was reviewed through discussion with chemistry personnel, review of applicable procedures and radio-
active release data and permits.
Procedures were established and implemented for liquid and gaseous wastes and reactor coolant as required by Technical Specifications.
Surveillance requirements are established in Procedure No. 1.8, " Master Surveillance Tracking ] Program."
j l The inspector determined that the licensee was meeting the requirements for sampling and analysis at the frequencies established in Technical Specification Tables 4.8-1 and 4.8-3.
l With regard to vendor supplied analyses of Sr-89, Sr-90, and Fe-55 l effluents, the inspector noted that the licensee controls vendor I activities through a sample QA program.
The inspector noted that the licensee had recently received copies of the applicable vendor
analysis procedures as well.
' The inspector reviewed the reactor coolant chemistry area against requirements in Technical Specification 3.6.B. " Coolant Chemistry" by review of procedures and records for total iodine, conductivity and chloride, and monthly isotopic analyses.
The licensee uses an offsite vendor for the monthly analysis requirements. The inspector reviewed analysis results through September, 1987 for total iodine and monthly isotopic analysis.
Within the scope of this review, no violations were found 8.2 Audits The inspector reviewed audits germane to the radioactive effluents control program required of Technical Specification 6.5.B and found them to be comprehensive in scope and generally adequate. Two audits were reviewed which covered 1986 and 1987.
Audit number
' 86-14 identified some findings which were adequately addressed in a timely manner and did not recur in the subsequent audit, 87-12.
Within the scope of this review, no violation were found.
! 8.3 Radiation Monitoring System l The inspector reviewed the gaseous and liquid effluent radiation monitoring system against surveillance requirements in Technical Specifications Tables 4.8-2 and 4.8-4, by review of procedures and records for calibrations, functional and channel checks through September, 1987 and found them satisfactory.
Specific reviews were made for systems and components in the offgas system, reactor building exhaust radiation monitor, control room ventilation intake process radiation monitor and liquid effluent radiation monitor.
There were no concerns in this area.
8.4 Offsite Dose Calculations The inspector reviewed release information and performed a spot check on the calculated doses in the liquid pathway for 1986 and the first two quarters of 1987, using the model equations in the Offsite Dose Calculation Manual (0DCM).
Due to the length of the extended shutdown, this was the only pathway of concern.
Spot check calculations performed by the inspector agreed with those of the licensee.
j l 8.5 Testing of Air Cleaning Systems The inspector reviewed the licensee's air filtration system testing j with regard to Technical Specification 3.7 requirements.
Surveillance in the control room for monthly systems operation and ! integrated SGTS run times were reviewed.
D0P and charcoal tests for methyl iodide removal efficiency for the control room environmental and SGTS systems were reviewed for 1985 and 1986.
The methyl iodide adsorption test is required once every five years.
{ The licensee uses an offsite contractor to perform the tests.
The j inspector reviewed test packages as follows: Date System Result . I i 3/7/85 Control Room Environmental Satisfactory Trains "A" and "B" 3/21/85 Standby Gas Treatment System Satisfactory Train "A", Banks 1 & 2 l 4/10/85 SGTS Train "B" Banks 1 & 2 Satisfactory j 10/9/86 Control Rcom Environmental Satisfactory Filter Bank "B"
-__-.____.______________J
l - Within the scope of this review, no violations were found.
i 8.6 Service Air System Contamination The licensee identified the existence of radioactive contamination ! in the station service air system. A routine surveillance on the Standby Liquid Control Tank revealed radioactive contamination which was traced back to the service air system which is used to sparge the tank. The source of ingress to the service air system has not been determined.
The service air system is a low pressure system used for operating valves and tools throughout the plant. A separate high pressure system is used for breathing air.
The licensee is pursuing resolution to this matter.
Within the scope of this review, no viola..sns were found.
9.0 Environmental Monitoring Program The purpose of this portion of the inspection is to review the licensee's k implementation of its radiological environmental monitoring program as described in Technical Specification Sections 7 and 8.
The inspection j included reviews of management controls, including program audits; sample analysis recoros; laboratory quality control results; sampling locations, field monitoring equipment operability, including meteorological moni-toring equipment; and calibrations of gas meters used in air sampling , equipment and of meteorological monitoring equipment.
! l 9.1 Management Controls The inspector reviewed the licensee's management controls for the Radiological Environmental Monitoring Program (REMP), including assignment of responsibility, program audits, and corrective actions for identified program inadequacies and problem areas in the program.
9.1.1 Assignment of Responsibility l The REMP is administered by a senior radiological engineer ' in the Radiological Engineering Group (REG). Although the i name of this group has changed since the last inspection j i in this area, its basic organization has not.
Some expansion has taken place, and there are plans to hire one j or two additional persons.
The REG Leader reports through { the Radiological Section Manager to the Manager of the j Nuclear 'herations Department.
Collection of i enviro we tal samples is performed by personnel in the l MechanicJ. & Chemical Test Group, General Test Division.
The supervisor of this group reports thrcugh the Head of
the General Test Division to the Manager M the Electrical l Engineering & Station Operations Department, j Environmental samples are analyzed by the yankee Atomic Environmental Laboratory, a division of the Yankee Atomic
-- _ __ - A '
N
)
I Electric Company. Oversight of the laboratory's activities is provided by a committee of representatives from the utilities that utilize its services.
i ! , The BECO representatives on this committee, known as the ' Laboratory Quality Control Audit Committee (LQCAC), are I the REG leader and the senior radiological engineer in ' charge of the REMP.
y I 9.1.2 Program Review and Audits The inspector reviewed audits of the REMP conducted by the . BEco Quality Assurance Department (QAD) in 1986 and 1987.
No deficiencies were identified in either audit. The 1986
' audit addressed follow-up of deficiencies identified in a previous audit. Two recommendations were made in the 1986 I audit, and a follow-up of these during the 1987 audit { indicated that both were implemented. One recommendation { was made during the 1987 audit, i The inspector also reviewed the LQCAC audits of the Yankee Atomic Environmental Laboratory (YAEL) conducted in August I 1986 and August 1987.
At the time of the last NRC j inspection of the REMP, some items from the 1985 LQCAC ] audit remained open.
The YAEL responses to these findings I were accepted in a meeting of the LQCAC held in April 1986.
9.2 Laboratory Quality Control The inspector reviewed selected records of quality control data submitted by YAEL to the licensee.
These data consist of both interlaboratory and intralaboratory QC samples. The interlaboratory program consists of participation in the USEPA Laboratory Intercom-parison Program. A large majority of QC samples were in agreement under the criteria used by YAEL and/or the EPA.
In cases where agreement was not obtained, a reanalysis of the sample and/or an investigation of the cause of the disagreement was undertaken.
The criteria used by YAEL were found to be acceptable.
9.3 Implementation of the Radiological Environmental Monitoring Program The inspector reviewed the licensee's commitments pertaining to its operational radiologici environmental monitoring program by dis-cussions with the licaisee, by review of reports, and by direct observation.
The inspector concluded that the program generally conforms to the licensee's technical specifications.
The inspector examined selected environmental monitoring stations including water sampling stations, air samplers for iodines and ~ particulate, and TLDs for direct radiation measurement. All L -- . - - - - - _ _ - - _ _ __
l
! equipment at these stations was operational at the time of this inspection.
The inspector also reviewed selected procedures for collection and handling of environmental samples.
The inspector reviewed Radiological Environmental Monitoring Program reports for 1985 and 1986.
These reports provided a comprehensive _i ' summary of the results of radiological environmental monitoring around the Pilgrim Nuclear Power Station, and met the technical specification reporting requirement (Section 6.9.C.2).
The licensee stated that plans have been made for improvement of the annual REMP report, and produced documentation to this effect. Among planned improvements are (1) revise the discussion of the monitoring results and the impact on the public of plant operations to make it . more understandable to the general public; (2) provide improved ! ! trending of environmental sample (including TLD) results.
The inspector stated that another area for improvement is in the reporting of the radiological impacts associated with the results of environmental samples.
Currently, the radiological impact of each sample medium, if any, is reported in the section devoted to that medium; no sunmary is provided.
A summary of such results,
especially if compared to the regulatory limits and to the results , of offsite dose calculations made on the basis of effluent data, . would improve the understanding of the actual radiological levels in i environmental samples.
' 9.4 Meteorological Motitoring The inspector examined the licensee's meteorological monitoring
system, including the primary and backup towers, the recorder charts in the equipment house at the backup tower, and the charts for the primary tower in the control room. There are no strip charts for - the backup tower in the control room.
l The wind speed, wind direction, and temperature sensors appeared to be operating properly at the time of this inspection. The inepector reviewed the log of instrument checks and calibrations that is maintained by the technicians who service the meteorological equipment.
This log indicated that the strip charts in the control room frequently required adjustment and/or maintenance, in large part due to an inability to run at a constant speed. The wind speed and direction charts were often found to be running slow by several minutes to several hours.
At the time of this inspection, the chart for the 220-foot wind speed and direction sensors was running approx-imately 9 hours fast.
The inspector noted that the guide holes on the left edge of the chart were not engaged in the sprocket. The licensee stated that replacement strip charts had been purchased and were available for installation, but that this had not been given a i high priority. The inspector stated that installation of these charts would be reviewed in a future inspection (293/87-35-03).
l _ - - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _.
. The licensee stated that a number of improvements to the meteoro-logical monitoring program.have been initiated, and that more are planned for the future.
In total, 21 proposed improvements were , listed in a memo from the senior radiological engineer.to the Radio-logical Engineering Group leader.
One of these planned improvements is the installation of new strip charts in the control room. Another is the use of a tower elevator to allow the sensors to be brought to ground level for service and calibration. This would permit calibra-tions to be performed under adverse weather conditions, especially during winter months. The inspector noted that calibrations were sometimes delayed due to ice on the tower and/or nonavailability-of an individual qualified to climb the tower. Also proposed and implemented was an audit of the meteorological data acquisition procedures.
Proposed but not yet implemented is an audit by a meteorology consultant of the existing program.
Some of the other j propcsed improvements are maintenance items such as new cables and structural improvements to the equipment houses..
' The inspector noted that the licensee's technical specifications contain no requirements for meteorological monitoring or for cali-bration of the meteorological equipment.
However, the licensee calibrates the equipment quarterly, and checks the calibration weekly.
l 9.5 Environmental TLD Program The inspector noted that the licensee has eliminated the environ-mental TLD system that was in use at the time of the.last inspection, and replaced it with a Panasonic system.
The. analysis of these TLDs is performed by YAEL.
The laboratory's quality' control program for this system was reviewed and found to be adequate.
The licensee had previously participated directly in the International-Environmental Dosimeter Intercomparison Program.
The licensee stated that it will
monitor the results of YAEL's participation in' this program in the future.
The NRC Direct Radiation Monitoring Network provides continuous measurements of the ambient radiation levels around commercial > nuclear power facilities (71 sites) in the United States.
Each site is monitored by arranging approximately 30 to 50 TLD stations in two concentric rings extending to about five miles from the facility. The monitoring results are published quarterly in NUREG-0837.
One of the purposes of this program is to serve as a basis of ! comparison with similar programs conducted by the. individual utilities that operate the commercial nuclear power plants.
Therefore, several NRC TLDs are collocated with licensee TLD l stations.
During this inspection, the monitoring results of collocated TLDs were compared and are listed in Table 6.
i >
- _ _ _ _ _ _ _ _ __ _-.
l The comparisons were made for the second and fourth quarter of 1986 and the first two quarters of 1987.
Only the second and fo;rth quarter data for 1986 were compared because net doses for the first i and third quarter NRC TLDs were unavailable (due to missing transit l ' control badges). The first column of data in Table 6 shows the comparison between the NRC TLD system and the licensee's old TLD system for the second quarter of 1986.
The remaining three columns show the comparison between the NRC TLD system and the licensee's new TLD system for the last quarter of 1986 and the first two quarters of 1987. The results of the comparisons between the NRC and the licensee are generally in agreement.
Some differences are to be expected, because some of the NRC stations are located as much j as a quarter mile from the corresponding licensee station.
I 9.6 Offsite Dose Assessment due to Hydrogen Water Chemistry Test j The inspector reviewed the licensee's assessment of the offsite doses that might result from the implementation of a hydrogen water chemistry (HWC) program at the Pilgrim site.
This program, when l implemented, is intended to mitigate and possibly eliminate inter- , , i l granual stress corrosion cracking in major plant piping systems by lowering the dissolved oxygen concentration in the primary coolant.
' This can be achieved by injecting hydrogen into the coolant.
This hydrogen addition also causes an increase in the radiation in the steam (primarily at the turbine) due to increased production of nitrogen-16, which has a very short half life of about 7 seconds.
l The increased levels of activity are present only while the plant i is operating.
l The licensee and its contractors performed extensive measurements and I calculations to assess the radiological impact of implementing an HWC program.
The inspector reviewed these measurements and calculations , with respect to offsite doses, to ensure that no regulatory limits
would be violated. The inspector found that even with hydrogen ' injection rates at the maximum proposed level, and using conservative t assumptions with respect to the exposure of persons outside the plant boundary, the regulatory limits (10 CFR 20 and 40 CFR 190) would not be exceeded.
It was further noted that the licensee had calculated the amount of shielding in the turbine building that would be required to bring offsite radiation levels down to near the levels that were i experienced in the past (i.e., without HWC), and that most of this shielding had been installed as of the time of this inspection.
Even if HWC were to be implemented at maximum injection rates, the actual offsite doses would probably be less than twice their previous levels with the additional shielding.
The licensee will perform additional tests to determine whether the installed shielding provides the expected reduction. At the time of this inspection, with the reactor shut down, measurements of background radiation levels were being made by personnel from YAEL. The licensee intends to make additional measurements during a subsequent test of HWC after the plant resumes operation.
L_-_-_____-______-____-_________-______________ _ _ _ _ _ _ _
i !
10.0 Exit Interview The inspectors met with the licensee representatives (denoted in Detail 1) I at the conclusion of the inspection on October 9, 1987.
The inspectors summarized the scope of the inspection and the findings.
The improvements in the radiological and nonradiological laboratory areas were noted as particularly significant.
I No information exempt from disclosure under 10 CFR 2.790 is discussed in this report.
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. TABLE 1 l WHOLE BODY COUNTING PHANTOM TEST RESULTS NRC Licensee Ratio Isotope Organ Known Value Result (Licensee:NRC) Co-60 Lung 62 8 69 9 1.1 0.2 Cs-137 Lung 91112 84 9 0.92 0.16 i Co-60 GI Tract 56 7 79 6 1.4 0.2 l Cs-137 GI Tract 82 11 121 10 1.5 0.2 i Note: All results in nanocuries.
,
_ _ _ - _ - _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _. . _ _. _
l ' .- l TABLE 2 RADI0 ACTIVITY CAPABILITY TEST RESULTS Sample Isotope NRC Value Licensee Value Comparison Results in Total Microcuries Charcoal Co-57 (4.00.2)E-2 3.7 0.2)E-2 Agreement Cartridge Co-60 (1.060.05)E-1 9.410.3)E-2 Agreement i Face Loaded Cd-109 (2.370.11) 1.9810.15) Agreement Detector 2 Cs-137 (1.0210.05)E-1 (9.10.5)E-2 Agreement Geometry 6 Charcoal Co-57 (4.010.2)E-2 (3.6 0.2)E-2 Agreement Cartridge Co-60 (1.06 0.05 E-1 (9.9 0.4)E-2 Agreement Face Loaded Cd-109 (2.37 0.11 (2.120.16) Agreement Detector 2 Cs-137 (1.02 0.05 E-1 (9.4 0.5)E-2 Agreement Geometry 11 Particulate Co-57 (3.8 0.2)E-2 (4.0 0.2)E-2 Agreement Filter Co-60 (9.90.5)E-2 /1.030.04)E-1 Agreement ' Detector 2 Cd-109 (2.2110.11) (2.2010.16) Agreement Geometry 2 Cs-137 (9.5 0.4)E-2 (9.8 0.5)E-1 Agreement Particulate Co-57 (3.8 0.2)E-2 (3.90.2)E-2 Agreement Filter Co-60 (9.910,5)E-2 (9.910.4)E-2 Agreement Detector 2 Cd-109 (2.210.11) (2.150.16) Agreement Geometry 8 Cs-137 (9.510.4)E-2 (9.6 0.5)E-2 Agreement . Charcoal Co-57 (3.9 0.2)E-2 (3.410.2)E-2 Agreement Cartridge Co-60 (1.020.05)E-1 9.310.3)E-2 Agreement Uniformly Cd-109 (2.29 0.11) 1.8310.14) Agreement Loaded Cs-137 (9.8 0.4)E-2 8.4 0.5)E-2 Agreement (Corrected Values) Detector 2 Geometry 11 Particulate Co-57 (3.8 0.2)E-2 4.1 0.2)E-2 Agreement Filter Co-60 (9.90.5)E-2 1.0410.04)E-1 Agreement Detector 1 Cd-109 (2.21 0.11) 2.410.2) Agreement Geometry 2 Cs-137 (9.5 0.4)E-2 (9.8 0.5)E-2 Agreement Charcoal Co-57 (4.010.2)E-2 (3.60.2)E-2 Agreement Cartridge Co-60 (1.0610.05)E-1 (9.310.3)E-2 Agreement Face Loaded Cd-109 (2.370.11) (2.1010.16) Agreement Detector 1 Cs-137 (1.020.05)E-1 (8.610,5)E-2 Agreement Geometry 6
_ _____- _-____________-_______- ___ __________ . _ _ _. __ _ _ _ _ - _ _ _ _ _ _ __
__ , .
l
TABLE 2 RADI0 ACTIVITY CAPABILITY TEST RESULTS
Sample E (kev) NRC Value Licensee Value Comparison i Results in Gammas per Second Simulated 186 160 8 13513 Agreement Offgas Vial 242 340120 272 4 Agreement ) Detector 2 295 850 40 690 6 Agreement Geometry 3 352 1640180 1341 9 Agreement >l Simulated 186 160 8 133 5 Agreement
Offgas Vial 242 340 20 27916 Agreement Detector 2 295 850 40 705110 Agreement Geometry 13 352 1640 80 1383 15 Agreement Simulated 186 160iB 12613 Agreement Offgas Vial 242 340 20 278 4 Agreement Detector 2 295 850140 732 6 Agreement Geometry 15 352 1640 80 1379 9 Agreement __________ _ __ -
- _ _____ .- . j TABLE 3 i CHEMICAL CAPABILITY TEST RESULTS Method Chemical NRC Licensee of Ratio Parameter Value Value Analysis (Lic/NRC) Comparison Results in Parts per Billion (ppb) Zinc 10314 104 4 DCP 1.01 0,06 Aqreement 28814 308 2 DCP 1.0710.02 Disagreement 480 6 503 4 DCP 1.05 0.02 Disagreement Fluoride 23.1t0.5 20.1 0.2 IC 0.8710.02 Disagreement 43.5 1.9 40.13 0.05 IC 0.92 0.04 Agreement 83.5 2.8 81.7 0.6 IC 0.98 0.03 Agreement Chloride 24.113.1 20.3 0.6 IC 0.84f0.11 Agreement 37.4 1.2 40.311.2 IC 1.08 0.05 Agreenent 80.5 2.2 81.0 1.0 IC 1.01 0.03 Agreement Sulfate 20.0 0.9 19.7 1.9 IC 0.98 0.09 Agreement 41.0 2.4 38 3 IC 0.9210.09 Agreement 80.8 3.0 77 2 IC 0.95 0.03 Agreement Chronium 153 9 155 4 DCP 1.01 0.07 Agreement 282 9 294.4 1.2 DCP 1.0410.03 Agreement 286 16 294.5 1.6 DCP 1.03 0.06 Agreement i Nickel 153 8 145.8e0.6 DCP 0.9510.05 Agreement ) 306 9 29014 DCf' O 95 0.03 Agreement j 306 8 295 2 DCP 0.96 0.03 Agreenent I Copper 140 7 141 2 OCP 1.01 0.05 Agreement 289 15 289 5 0CP 1.00*0.06 Agreement 290r12 287.2 0.8 DCP 0.99 0.04 Agreement Iron 48.913.5 46.1 0.2 DCP 0.9410.07 Agreement 95.5 3.4 97.1 1.8 DCP 1.02 0.04 Agreement 147 4.2 142.0 1.0 DCP 0.97 0.03 Agreement Silica 54.3 5.6 51.8 1.2 SP 0.95i0.10 Agreement 109 7 99.2 6.3 SP 0.91 0.08 Agreement 160 5 151.3 2.1 SP 0.9510.03 Agreement - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ - _ _ _ - - ___ -_ __-_______-__ ____ _ _ _.
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, . .1 TABLE 3 CHEMICAL CAPABILITY TEST RESULTS l Method i Chemical NRC Licensee of Ratio Parameter Value _V,alue Analysis (Lic/NRC) Comparison Results in Parts per Million (ppm) f Baron 1000 10 983121 Titration 0.9810.02 Agreement 3024 46 2970 10 Titration 0.982 0.015 Agreement ! 4947 61 4957 21 Titration 1.002 0.013 Agreement i NOTE: OCP = direct current plasma spectrometry IC = ion chromatography SP = uv-vis spectrometry i ! I I I
l l l I i
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TABLE 4 PASS CAPABILITY TEST RESULTS
Chemical NRC Licensee Method Parameter Value Value of Analysis 10% Pesults in Parts per Million (ppm) Boron 1.000 0.010 1.017 0.015 DCP Yes ] 3.02 0.05 3.04 0.04 DCP Yes . ' 4.95 0.06 4.97 0.11 DCP Yes 1 Boron 1.000 0.010 0.80 0.02 SP No 3.02 0.05 2.91 0.08 SP Yes 4.95 0.06 5.0 0.2 SP Yes 2Boron 1.000 0.010 0.913 0.015 SP Yes 3.02 0.05 2.90 0.08 SP Yes 4.95 0.06 4.8 0.2 SP Yes Boron 1.00010.010 1.0110.02 SP Yes 3.02 0.05 2.96 0.08 SP Yes l 4.9510.06 4.9 0.2 SP Yes l l NOTES: 1.
0-15 ppm calibration curve results l 2.
0-8 ppm 3 point calibration curve results 3.
0-8 ppm 5 point calibration curve results '
l l l l l __ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
I l J ,-
TABLE 5 . SPIKED SAMPLE RESULTS Chemical NRC Licensee i Sample Parameter Value Value Results in Parts per Billion (ppb) 86W Iron 471 508 Makeuo Water Copper 533 513 Nickel 499 487 I Chromium 436 507 i 86V Chloride 26.8 24.2 Reactor Water Fluoride 26.2 47.3 Sulfate 29.4 26.4 Results in Parts per Million (ppm) Standby Liquid Baron 24,850 22,363 Control l l l - _ _ - _ _ _ _ _ -
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ ,- .* I ) TABLE 6 INTERCOMPARIS0N OF ENVIRONMENTAL TLD MONITORING RESULTS (mR/std. quarter i random uncertainty) Station Number Monitoring Period i 1986 1987 i NRC Licensee 2nd 4th 1st 2nd No.
No.
Otr (1) Otr(2) Qtr(2) Qtr (2) !
26.8 i 0.8 15.7 0.9 17.1 0.9 14.3 0.7 PL 25.0 5.9 17.3 1.2 13.5 1.0 18.6 2.4
22.3 0.7 14.6 i 0.9 14.8 0.9 (3) PA 32.6 2 7.7 18.1 i 1.1 17.9 1 1.3 17.8 1 0.8
22.8 0.7 15.7 0.9 15.8 0.9 16.3 0.8 BD 26.8 6.3 17.5 1.1 15.4 1.0 19.3 1 1.4
21.6 0.7 14.2 0.9 14.3 0.9 15.2 1 0.8
EM 26.2 t 6.2 17.6 2.1 13.8 1.1 18.6 2.0
19.7 0.6 12.0 0.8 (3) 13.5 0.7 MP 32.3 7.6 17.8 1 1.6 (3) 17.9 1.5
21.2 0.7 12.5 1 0.9 13.0 0.8 14.3 1 0.7 - ' WH 28.7 6.8 16.7 1.4 13.7 1.1 18.2 2.2
(3) 20.7 1.1 14.5 1 0.9 16.1 0.8 MS 29.5 7.0 20.4 1.9 16.8 1.5 21.0 1 3.1
18.9 1 0.6 13.7 i 0.9 14.7 0.9 14.6 0.7 PC 19.2 4.6 11.8 0.9 11.7 1 0.6 12.4 1.1
18.7 0.6 12.8 0.9 14.9 0.9 12.2 1 0.7 MB 24.3 2 5.7 16.4 1.4 16.2 1.3 16.7 2.0
22.1 0.7 14.6 * 0.9 16.5 0.9 14.9 0.8 NP 32.6 7.7 18.1 2 1.4 15.7 1.1 17.5 1 1.1 NOTES: (1) Licensee's old TLD system (2) Licensee's new TLD system (3) Missing or damaged dosimeter; no comparison possible _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ -
4'
ATTACHMENT 1 CRITERIA FOR COMPARING ANALYTICAL MEASUREMENTS-TABLE 2 This attachment provides criteria for comparing results of capability tests and verification measurements.
The criteria are based on an empirical relationship which combines prior experience and the accuracy needs of this program.
In these critaria, the judgement limits are variable in relation to the comparison of the NRC Reference Laboratory's value to its associated uncert ainty. As that ratic, referred to in this program as " Resolution," increases the acceptability of a licensee's measurement should be more
selective.
Conversely. poorer agreement must be considered acceptable as i the resolution decreases.
'
Resolution Ratio for Agreement 2
0.4 - 2.5 ,
4-7 0.5 - 2.0 8 - 15 0.6 - 1.66 ' 16 - 50 0.75 - 1.33 51 - 200 0.80 - 1.25 200 0.85 - 1.18 , I Resolution = (NRC Reference Value/ Reference Value Uncertainty) d I l 2 Ratio = (License Value/NRC Reference Value) l
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R l j
i l
. i i ATTACHMENT 2 { CRITERIA FOR COMPARING ANALYTICAL MEASUREMENTS-TABLE 3 l l This attachment provides criteria for comparing results of capability tests.
l In these criteria the judgement limits are based on the uncertainty of the i ratio of the licensee's value to the NRC value.
The following steps are
performed: j (1) the ratio of the licensee's value to the NRC value is computed I (ratio = Licensee Value)I I NRC Value l (2) the uncertainty of the ratio is propagated.1 j If the absolute value of one minus the ratio is less than or equal to ,
twice the ratio uncertainty, the results are in agreement.
(ll-ratio 2 uncertainty)
2
2 I t f=f+[S S S ' 1 7, , then I(From: Bevington, P.R., Data Reduction and Error Analysis for the Physical Sciences, McGraw-Hill, New York, 1969) ! l l l r -___.__.___._._______m }}