IR 05000321/1990024
| ML20029C046 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/05/1991 |
| From: | Elliott M, Gloersen W, Seymour D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20029C036 | List: |
| References | |
| 50-321-90-24, 50-366-90-24, NUDOCS 9103250163 | |
| Download: ML20029C046 (45) | |
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2tso UNITED ST ATES.
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NUCLEAR REGULATORY COMMISSION u
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MAR (j 6 1991 i
Report Hos.: 50'321'/90-24-and 50-366/901 ~4 I
Licensee: Georgia; Power. Company
P.O. Box 1295.
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Birmingham, AL 35201
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A Docket Nos.: 50-321-and 50-366--
License Nos.: DPR-57 and NPF-5 l
Facility Name: 1 Hatch 1 and 2 f
- Inspection Conducted-cember. 10-14, 1990
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Approved by:
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Facilities; Radiation Protection Section -
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Radiological. Protection and Emergency
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Preparedness Branch
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i SUMMAR_Y.
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'4 Scope:!
- This was a-special announced assessment-of the111censee's program to maintain T
. occupational radiation exposures as low as reasonably-achievable-(ALARA).
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a During the period of December 10-14, 1990, a special team assessment was
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1 conducted by.Sthe' NRC to evaluate the licensee's efforts ifor maintaining.
- occupational radiation doses as low as. reasonably -achievable (ALARA) : The assessment included a review of the causes of the past and current;high radiation doses; an evaluation of the licensee's current organizationf and.
_ program for keeping radiation doses ALARA; a review of past and current licensee initiatives to bring the radiation doses to within industry norms; and-M an evaluation. of the licensee management's awareness of, involvement in, and support for-the ALARA program. In addition, on January 24, 1991, the licensee 9103250163 910306 PDR ADOCK 05000321 g
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and General Electric presented their assessment of the hydrogen water chemistry impacts in a meeting in NRC Headquarters.
The assessment team found an adequate level of piant management awareness of and support for the dose reduction program. The licensee nas taken a number of initiatives, 'ncluding the budgeting for the chemical decontamination of the recirculation pipes and an aggressive program to minimize the number of personnel contimination events.
Continued management support will be necessary as problems artse which may conflict with the dose reduction goals.
Program strengths and areas for improvement were identified during the assessment and are sunnarized as follows:
Strengths Addition 'of an ALARA module in 1987-1989 in the GET and continuing
training programs (Paragraph 6).
Outstanding ) personnel contamination reduction during the last two years
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(Paragraph-7.
Good reduction in percentage of rework and applicatiorr3 of lessons learned (Paragraph 11).
Ensuring an effective measurement system is implemented-to determine the
degree of success with regard to goals and objectives by inclusion of dose goal performance as part of supervisory personnel annual performance review (Paragraph-7).
GeneralworkerandmanagementknowledgeofALARAconcepts(Paragraph 6).
- Proactive approach to - source term reduction and the dose impacts
- mitigating from the hydrogen water chemistry program (Paragraph 10).
Improvement Areas e
Inconsistent use: of tool packaging for maintenance work orders (Paragraph 4.d(1)(c)).
Better review of design change requests for maintenance items
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-(Paragraph _4.c).
Improved: coordination of job-activities in the. drywell (Paragraph 4.d(1)(c)).
Improved dose. goal awareness for plant employees (Paragraph 4.c)
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Minimal use of the ALARA suggestion program (Paragraph 4.c)
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Improved performance of the plant ALARA Review Con'mittee (Paragraph 4.c)
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j-Improved corrective action of QA Audit findings (Daragraph 11.a).
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REPORl DETAILS 1.
Persons Contacted Licensee Employees
- J. Betsill, Operation Unit Superintendent
- K. Breitenbach, Supervisor, Engineering Support R. Bryant, Plant Chemist
- P. Fornel, Manager, Maintenance
- 0. Fraser, SAER Site Supervisor
- G. Goode, Manager, Engineering Support
- J. Hammonds, Regulatory Compliance Supervisor
- D Hopper, Project Supervisor, HP/ Chemistry (Corporate)
- R. King, Engineering Supervisor
- W. Kirkley, Engineering Supervisor, HP/ Chemistry
- M. Link, HP Supervisor
- B. Manning, Nuclear Specialist, Quality Assurance
- P. Moxley, HP Specialist
- J. Reddich, HP Support Supervisor
- R. Reddick, Emergency Preparedness Coordinator
- M. Rigsby, Plant Health Physicist
- J. Robertson, Supervisor, Engineering Support
- D. Smith, HP Superintendent
- L. Sumner, General Manager
- R. Zavadoski, Manager HP/ Chemistry Other licensee employees contacted during this inspection included craftsman, engineers, operators, mechanics, technicians, and administrative personnel.
Other Organizations C. Brossier, Site Services Manager, General Electric Nuclear Regulatory Concission
- J. Potter, Section Chief, Facilities and Radiction Protection Section, Region II
- Attended exit meeting 2.
Program for Maintaining Exposures As Low As Reasonably Achievable (ALARA)
(83728)
a.
Regulatory Guidance 10 CFR 20.lc states that persons engaged in activities under licenses issued by the NRC should make every reasonable effort to maintain
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- radiation exposures a low as reasonably achievable.
The recommended elements of an ALARA program are contained in Regulatory Guide 8.8, Information Relevant to Ensuring that Occupational Radiation Exposure at Nuclear Power Stations will be ALARA, and Regulatory Guide 8.10, Operating Philosophy for Maintaining Occupational Radiation Exposures ALARA.
Regulatory Guides (RG) 8.8 ard 8.10 provide infornation relevant to attaining goals and objectives for planning and operating light water reactors and provide general philosophy acceptable to the NRC as a necessary basis for a program of maintaining occupational exposures as low as reasonably achievable (ALARA).
b.
Objective The objective of this assessment was to evaluate the effectiveness of actions the licensee has taken or are being taken to control collective dose and to determine whether the licensee is meeting commitments to make reasonable efforts to ensure that occupational radiation exposures are maintained as low as is reasonably achievable (ALARA).
This inspection was conducted at the E. I. Hatch facility because of the historically higher than industry average collective dose for personnel ccupied with a significant collective dose this calendar year.
Th;s assessment examit.ed the licensee's ALARA program and organization, corporate involvement, ALARA training, management goals, procedure implementation, planning and scheduling, ALARA initiatives and operational practices, and licensee assessments and self evaluations.
3.
Background (83750)
During the six year period from 1984 thru 1989 Hatch exceeded the U.S. BWR reactor national average collective dose per unit in three of the six years as reported in NUREG-0713.
As of DecemLer 8, 1990, the collective dose per unit was 750 person-rem.
Table 1 lists the Hatch performance measured against the NUREG-0713 per unit average collective dose in U.S.
BWRs for the period 1984 thru 1990.
Major work contributing to the collective doses in 1984 included recirculation pipe replacement on Unit 2 and activities associated with that job.
In 1984, during the Unit 1 outage (welding, UT, PT, NDT, and decontamination), the welding portion of the recirculation pipe weld overlays accounted for approximately 78 person-rem and weld testing accounted for another 58 person-rem.
Table 2 lists the major jobs and the collective dose accumulated and the actual dose versus estimated dose for the Unit 1 outage.
The estimated person-hours versus the actual hours expended is also tabulated.
Significant growth in actual person-heurs expended is shown in conduit installation, recirculation pipe welding and snubbers / torus jobs.
Major work contributing to the collective doses in 1986 are tabulated ir l
Table 3.
Both Unit 1 and Unit 2 had significant outage work performed.
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Table 4 shows major work activities during the 1988 Unit 1 and Unit 2 outages.
Major jobs contributing significant collective dose included drywell insulation, drywell painting, pipe replacement for the 631 system, and RWCU heat exchanger replacement.
In the 1990 Unit 1 outage, collective doses for significant jobs are shown in Table 5.
This outage occurred after an extended run of 424 days and included hydrogen injection water chemistry, and zinc injection.
Admiralty brass condenser tubes were replaced on both units, and Zinc injection has been implemented on Unit I since the outage.
Dose rates in the drywell at shutdown were found to be approximately 3 to 5 tines greater than expected.
A hydrogen injection test was conducted December 6,1993, on Unit I and the average value for the A, B, C, and 0 Main Steam Line Radiation Monitors versus hydrogen injection rate is presented in Table 6.
Background radiation dose rates from the production of N-16 during operation showed an increase as the hydrogen injection rate was increase <'
Radiation monitor surveys inside and outside the plant showed increases.
The dose rate increase derived from the hydrogen addition and N-16 production ceased at shutdown.
The licensee has recognized that the dry well source term has been a major contributor to the Unit 1 outage work.
In order to reduce the dose rates to the maintenance workers during outages, the licensee has established a task force and budgeted monies to perform chemical cleaning.
The present schedule for Unit 1 shows this chemical cleaning taking place during the neyt Unit i refueling outage.
At the present time, Unit 2 was not being operated with hydrogen injection.
Chemical cleaning of Unit I was expected to reduce the dry well source term 5 to 10 times below the current dose rate.
4.
ALARA Program and Organization a.
ALARA Program Section C.I.a of RG 8.8 specifies the establishment of an ALARA program.
The licensee's ALARA program w.s described in Administrative Control Procedure 60AC-HPX-009-05, "ALARA Program," Revision 5, October 2, 1990 and was a plant wide procedure with responsibi.lities assigned to the various departmental managers including the General Manager's office, superintendents, supervisors. foremen, and all plant personnel. The Health Physics and Chemistry Department was primarily responsible for establishing, implementing, and supporting the ALARA program. In addition, the HP/ Chemistry Department had the authority to halt cny activity which had been deemed radiologically unsafe, or was outside the parameters discussed in ALARA pre-job planning.
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60AC-HPX-009-0S also utilized Regulatory Guides 8.8 and 8.10 as developmental references.
b.
Organization and Staffing Section C.1.b. of RG 8.8 specifies the organization, pemnnel, and responsibilities of an ALARA program.
The licensee's ALARA support organization consisted of: (1) a Health Physics - Specialist with responsibilities -for radiological work permits (RWPs), RWP man-hours and man-rem budgets, ALARA work i
histories, shielding program, hydrogen injection surveys, and pre-and post-job ALARA reviews; (2) the Plant Health Physicist with responsibilities for the Plant " ARA Review Coneittee (PARC),
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L collective dose tracking and tiendir.9, initial development of and-(3)partment collective dose goals, and personnel contaminations; site /de L
two'HP Technicians with rotational assignments in the ALARA group. The rotational duty was typically-for one year. In addition, the HP/ Chemistry Depe. ment assigned an individual to function as a
. work controls representative. During outages, additional HP j
technicians would be added to supplement the ALARA operations staff._
The overall quality and experience of the ALARA -personnel appeared
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generally good. The HP Specialist and Plant HP had several years of
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BWR plant experience.
Licensee personnel interviewed indicated that
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existing staff size-to manage the operational ALARA program and the supplemented staff to manage the outage ALARA program were sufficient to implement the program.
It was noted in Administrative Control t
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L Procedure 60AC-HPX-009-0S that all jobs which were expected to have a
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collective dose greater than one person-rem will have a detailed ALARA review performed and ALARA briefings required. In addition, Ljobs expected to receive less than 10 person-rem iequired an HP For.eman's approval; jobs expected to receive greater that or equal to-10 person-rem but less than 25 person-rem required an HP Supervisor's approval;~and. jobs which were expected to be greater than or equal to-25 person-rem required approval by the PARC as deemed necessary by the PARC Chairman.
The inwectors also veriH.d' that specific responsibilities for l
conducting the ALARA program have been assigned to the following positions and/or organizations via Administrative Control Procedure 60AC-HPX-009-05: (1) General Manager and Assistant General Managers; (2) Health Physics ~and Chemistry Department;._(3) Manager of Health Physics and Chemistry; (4) other <lepartr
.al managers; (5) Training
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_ and Emergency Preparedness Departme % (6) Plant ALARA Review
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Coneittee; (7) Superintendents, Supervisors, and Foremen; and (8) all plant personnel.
-In addition, the inspectors verified that an adequate written management policy as recommended in RG 8.8, Section C.1.a. was issued to cover the ALARA program. The ALARA program responsibilities were L
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- specified in a n,emo to all plant personnel from the General Manager dated December 5, 1990. This memo was in essence a revision to a very similar nemo dated Novenber 14, 1988. The revised meno basi ally deleted the reference that the PARC was responsible for interfacing with the Corporate Radiation Safety Group. The inspectors also noted that an explicit written endorsenent of the ALARA program from corporate managen ent was docun ented in a policy statement (N0P-07-100) dated July 28. 1986.
c.
ALARA Program Awareness, Support and Incentives Section C.1.f of RG 8.10 specif rs that n.odifications to operating and naintenance procedures and to plant equipment and facilities should be implementeo where practicable.
Section C.1.b of RG 8.8 describes the responsibilities 6nd authorities an ALARA Committee.
Support of ALARA efforts was evident in sore area and appeared to be continuing in that the licensee has budgeted funds to perform a chemical decontamination of the Unit I recirculation piping during the next refueling outage scheduled for Fall 1991. Further discussions of other initiatives undertaken by the station are discussed in Section 10.
Support of the ALARA program n other areas, including the PARC activities, management support of and worker participation in the ALARA program, and worker awareness was not as evident. In the paragraphs below, these items are discussed in greater detail.
The inspectors reviewed and discussed ; % activities of the PARC with licensee representatives and reviewet. $RC meeting minutes from May 1988 through November 1990. The PARC membership was formed in 1988 with personnel from key station departments, including: Health Physics and Chemistry, Plant Engineering, Mcintenance, Operations, Operations and Maintenance Planning, Instrument and Controls, Outage and Planning, Plant Engineering, and Building and Grounds. The PARC Committee Chairman was the Manager of Health Physics and Chemistry.
PARC meetings were scheduled on an as needed basis by the chairperson. A m%imum of five days notice prior to each meeting was required. The PARC findings and recommendations were generally made known to the Plant General Manager. The PARC was established to perform reviews and evaluations of changes, events, and work practices; review the station's collective dose goal; review jobs whose collective dose would be equal to or exceed 26 person-rem; review exposure reports, personnel contamination reports, and radiological deficiency reports; pursue, review, and respond to employee ALARA suggestions; and ensure that corporate ALARA radiation protection policies were implemented.
Based on a review of the PARC neeting minutes, the inspectors made the following observations:
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non-regular scheduling.
Pest attendance of the Operations Department was poor in that a representative attended only 20 percent of the time. Several other key departments attended the meetings between 50 and 60 percent of the time. A representative from the Corporate ALARA group was not included in the membership.
PARC meeting minutes did not,ndicate that the fundamental
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elements in the PARC charter vere routinely discussed, some examples included: ALARA sugges+1ons, collective dose goals, and radiological deficiency repoi+.s. The inspectors observed
that the PARC was more involved in olanning jobs expected to exceed 25 person-rem and with work com dination problems identified during the outages.
Specific methods to reduce dose for jobs exceeding their
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collective dose budget were not always discussed however the majority of requests for additional dose allowance for jobs were approved by the PARC..
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The PARC developed an action item list to track problems experienced during outages that were identified in post outage debriefings. These items were either assigned to all departments or a specific department for action with no action _due date specified.
The inspectors also assessed staff awareness of the ALARA program by conducting interviews with managers and station personnel, reviewing station records, and by direct observation. One area of_ management commitment' to-the ALARA program was the incorporation of station collective dose goals to the departmental. managers' and supervisory personnel performance appraisals. - However, during interviews, some departmental managers and supervisory personnel were not aware of their respective department's dose goal. In addition, there was a rather inactive ALARA suggestion _ program. There were eight ALARA suggestions submitted in 1W8.and five in 1989. It was difficult to determine the number of suggestions received in 1990 since items from post job debriefings were added to the PARC action item list and counted as ALARA suggestions. Most of the PARC action items were work coordination problems and 'were not quantified with _ respect to dose. The inspectors also noted '
that the station did not have an incentive program to create interest and competition for submitting good quality technical dose-saving ideas.
During interviews,.most managers and workers were unable to identify the locations of the three ALARA suggestion boxes.
During tours of the facility, the inspectors did not observe the use of ALARA posters or signs to enhance the interest and awareness for the ALARA
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Based on the above review, the following improvement items were ident;fied regarding the licensee's ALARA program:
(1)
Improvement of the Plant ALARA Review Contrittee functions and attendance.
(2)
Improvement of the ALARA awareness of station personnel by consideration of the following:
Improvement of the awareness of departmental dose goals.
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Use of ALARA posters and signs to enhance interest and
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awareness for the ALARA program.
Improsenent of the awareness of ALARA suggestion box
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locations and cor sideration of the use of an incentive program to elicit ALARA suggestions, d.
Interviews (1)
Empicyee Discussions Discussions were held with licensee employees to assess their knowledge, involvenent, and perspective of the utility's ALARA program, including the employees' knowledge of ALARA goals, concepts, policies, and procedure documents; -individual responsibilities, personnel doses and personnel dose limits; the employee's involvement in special ALARA training, conmunication with co-workers and supervision, and participation in the ALARA suggestion program; and the employees perspective on how to improve the ALARA program, what events or condition have caused increased personnel doses, and on what events or conditions had-helped reduce personnel doses.
(a)
Employees All employees interviewed entered the radiological controlled areas (RCAs) on a daily to weekly basis depending on plant conditions.
(b)
Knowledge of ALARA Program Each of the employees interviewed was familiar with the basic At ARA concepts taught in the General Employee Training (GET) program and knew that they had a basic responsibility for implementing the utility's ALARA program by performing tasks in a manner cnnsistent with the utility's ALARA policy.
Ali of the employees knew their current radiation dose and their dose limit.
Most employees were aware of where the ALARA requirements originated and what documents described the ALARA program
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objectives.
With the exception of the health physics staff, none of the employees interviewed knew that each of their departments had an ALARA goal, and generally were not aware of the goal that was established.
Many employees interviewed believed that the health physics staff was primarily responsible for ALARA.
(c) ALARA Program Involvement The employees interviewed had received basic ALARA training provided in the GET course.
A large number of those interviewed had received some informal ALARA training on jobs requiring ALARA pre-job planning and on-the-job training.
Employees reported frequent discussions of ALARA objectives on major jobs during outages with co-workers and supervisors.
The employees also reported good communication with the ALARA and HP staffs.
None of the employees interviewed had participated in the formal ALARA suggestion program.
Other employees reported that they had made suggestions to supervisors informally and had not used the formal ALARA suggestion program.
Several of the employees mentioned the inconsistent use of tool packaging for maintenance work orders.
Pre-packaging of tools for various job was not a formal program.
The current method depends on the individuals initiative to ensure that the tool lists are updated and maintained. Job histories were not always complete and therefore did not lend themselves to tool prepackaging.
Improvement of job coordination activities inside. the drywell to better schedule respirator use and man loading was mentioned by many of the employees interviewed.
The inspector informed licensee representatives that RG 8.8, Section C.3.a, reviewed that station personnel should have the benefit of preparations and plans v,hile personnel are performing services in radiation areas.
(d) Perspective Several of the employees had suggestions on how the ALARA program could be improved.
Suggestions included better planning and scheduling of work to ensure that appropriate equipment and tools were readily available to perform tasks expeditiously and installing more easily removable shielding to minimize the time spent removing and installing the same shielding package.
The majority of employees had an opinion on what had contributed to decreases and increases in personnel doses.
Employees believed that the following actions had contributed to dose
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reductions:
use of temporary shielding in the drywell and decontamination of contaminated areas within the RCA.
Employees believed that the following actions had contributed to increases in personnel doses:
hydrogen water chemistry, drywell recirculation pipes, and valvet, pumps and nozzles.
(2) Management Discussions Discussions were held with licensee managers and supervisors to assess their knowledge of the utility's ALARA program including the manager's or supervisor's knowledge of ALARA goals, concepts, policies and procedure documents, individual responsibilities, personnel dose, and personnel dose limits; the manager /supervior's involvement in special ALARA training, connunications with co-workers and supervision, and participation in the ALARA suggestion program; and the manager or supervisor's perspective on how to improve the ALARA program, what events or conditions have caused increased personnel doses and what events or conditions have helped reduce personnel radiation doses.
(a) Managers and Supervisors Most of the individuals interviewed entered the RCAs on a weekly to monthly basis depending on plant conditions.
(b)
Knowledge of ALARA Program Each of the individuals interviewed was familiar with the basic ALARA concepts taught in the GET program and knew that they had a -basic responsibility for implementing the utility's ALARA program by performing a task in a manner consistent with the utility's ALARA policy.
In general, the managers and supervisors interviewed were knowledgeable of the current radiation exposure, however, some of the managers were not familiar with the dose goals for their departments.
The managers and supervisors understood where the ALARA requirements originated and what corporate and plant documents described the ALARA program-objectives.
(c) ALARA Program Involvement The managers and supervisors interviewed had received the basic ALARA training provided in the GET course.
(d)
Perspective Most managers and supervisors interviewed had suggestions on how the ALARA program could be improved.
The suggestions included purchasing additional temporary shielding, recommendations for better scheduling and
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planning of work, ensuring that appropriate equipn;ent and tools were readily available, and continuing to increase the awareness of.the ALARA concept to all levels of plant personnel including the use of ALARA suggestion boxes.
These methods could be implemented through GET retraining, departmental training, and non-licensed training.
The majority of managers and supervisors had opinions on what had contributed to the increases in personnel exposures.
Individual managers and supervisors interviewed believed that the following actions had contributed to exposure reductions:
reduction in the number of personnel contaminations, use of temporary shielding; reduced work activities in high radiation areas; and reduction of contaminated areas within the RCA.
Individual managers and supervisors interviewed believed that the following actions had contributed to increases in personnel exposures:
poor coordination of job activities in the _drywell, hydrogen.
water chemistry: and increased maintenance activities in j
the drywell.
5.
Corporate involvement
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Section C.1.b of RG 8.8 specifies the responsibilities of a corporate ALARA program.
The corporate support for radiological safety consisted of a Project Supervisor of Radiation Protection and-Chemistry
'(PSRPC). The PSRPC reported directly to the Manager of Environmental Services. The Manager of Environmer, cal Services reported directly to the Vice President of Technical Services.
The' corporate support for the licensee's. ALARA prgram resided solely with one individual, - the PSRPC.: The PSRPC was also responsible for suptrvising a-technical staff of three nuclear-a specialists and ong engineer (at -the time of this inspection, there were two vacancies) to accomplish the following:
Provide technical support to the Projects in radiaticn
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protection, ALARA, radioactive - waste, radiochemistry, radiological environmental monitoring, and other related ~
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Identify, track, and provide technical evaluations of
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selected regulatory / industry issues related to assigned functional. areas.
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Provide the technical interface with industry groups such as EEI and EPRI.
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Provide technical and task support of ALARA programs, resolution of INP0/NRC concerns, special technical reviews, and outage activities as requested by the Project or Technical Services Management.
Upon request, provide reviews of plant performance data
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and support plant staff in identifying trends and resolving identified issues.
At the Project request, provide management and
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technical interface for offsite radiological monitoring activities by the GPC Central Lab.
At the Project request, provide preparation and/or
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review of routine radiological environmental reports required by the State and NRC.
It appeared that the corporate ALARA organization served as a support service and apparently was not involved proactively with reviewing relevant dose reduction research, practices, and modifications performed in the nuclear industry and disseminating this information to the appropriate individuals within the organi?ation. Additionally, the PARC did not have a permanent member from the corporate office attending the meetings.
The inspectors also noted that the corporate office performed an ALARA assessment several weeks before this inspection. Howe"er, the assessment was informal, not documented, and therefore did not receive the proper management review.
Overall, the corporate support of the ALARA program appeared broad in scope but only marginally effective because it consisted of only limited involvement by one individual.
6.
ALARA Training Sections 1.c of RG 8.8 and 1.d of RG 8.10 specify the elements an ALARA training and instruction program.
The inspectors reviewed selected licensee training programs regarding presentation and implementation of ALARA policies and procedures for routine and special work activities.
Information was collected by interviews with licensee personnel; procedure and policy reviews; review of instructor lesson plans and training course materials; and tours of licensee training facilities.
The inspectors noted management's commitment to quality training as indicated by the modern, well equipped laboratories, equipment, and facilities.
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General Employee Training (GET)
The GET program was accredited by the Institute of Nuclear Power Operations (INP0).
The GET lesson plans and training course materials were revised to include ALARA in January 1989.
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GET program consisted of four types of GET:
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Initial GET was required for all employees that have never completed an INPO certified GET program or have not completed one in the past three years.
Initial GET provided three days of training which included ALARA topics.
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Exemption GET was provided to those employees that have completed an INP0 certified GET program within the past three years.
Exemption GET involved one day of training which included.ALARA topics, 3.
Annual Requal GET was required for all employees on an annual basis and also included emphasis on ALARA.
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Plant ALARA Review Ccmmittee (PARC). Training was required for r
all PARC members init Ally.
This training was provided by the GET staff and contained topics which include ALARA program responsibilities and goals, pre-job planning, ALARA practices during job performance, post job reviews, facility design, dose-projections, and cost benefit analysis, b.
Radiation Worker Training Interviews with the licensees training representatives and a review of selected records indicated that all radiation-workers must
. complete GET.
Radiation workers are then trained in their respective disciplinary areas which include topics on ALARA specific to their-type of work.
The disciplinary areas also included ALARA in their continuing training programs in addition to that given in the Annual Requal GET.-
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Health Physics Technicians y
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Interviews with -licensee training representatives and Health Physics management and ' staff indicated that in-addition to GET, HP technicians received six weeks-of academic training and six weeks of-
" Job Coverage" training.
After successful. completion of class-work the technician participated in on-the-job training (0JT).
0JT consisted of.several modules representing different HP areas of which
'the trainee must demonstrate competence in before being allowed to work independently in the area represented by the module.
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supervisor or foreman must observe the traince's performance and then
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may sign the authorization for the technician to be certified under that module.
All 0JT modules contcined practices, procedures and policies that included acceptable ALARA practices.
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The academic and " Job Coverage" class-work also contained two weeks each of systems treining to provide the technician with the knowledge to make decisions on working environments to include ALARA.
d.
Corporate Personnel Interviews with corporate HP representatives indicated that no formal ALARA training was available to them.
However, Corporate HP Staff did attend various professional development activities such as Westinghouse's REM Seminar and EPRI and INPO workshops, e.
ALARA Coordinator Training Interviews with licensee rep; u entatives indicated that no one individual had the responsibility of the "ALARA Coordinator",
The inspector noted, however, that the individuals in the HP department involved with the ALARA program implementation held Bachelor of Science degrees or higher and had completed all Georgia Power required training courses and did attend professional development meetings and seminars, f.
Training Program Instructors The training program pertaining to ALARA utilized two HP instructors.
They held A.S. and B.S. degrees and had 10 to 15 years of plant experience.
Both received training through outside contractor courses, professional meetings and seminars and working as plant HP technicians 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> per year.
Instructors were evaluated on an annual basis for technical and instructional abilities.
g.
Training Feedback Mechanisms The Training Advisory Conmittee provided technical feedback and guidance to the training programs for all disciplines and met at least quarterly.
The Plant Training Review Board provided guidance and recommendations for major changes to the training program.
Members included the Plant General Manager, Department Manager representatives and the Training Manager.
The Board met approximately once per month.
Instructors performed actual HP technician duties for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> per year for practical experience. This experience has proven to be a good feedback mechanism.
The training department also reviewed all radiological deficiency reports for root cause and applicability to training.
Quarterly radiological deficiency trends were also reviewed for training applicability.
The Program Developtrent section of the training department submitted items such as industry events, NRC Information Notices, and other items that may be applicable for training to the training instructor management consistent with RG 8.8 guidance.
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Management Goals b
Section 1.b.(2) of RG 8.8 describes the selection of specific goals and objectives for the station.
ALARA goal setting at Hatch was evaluated during review of applicable docuntentat.on and interviews with plant personnel.
Areas examined incluced historical dose goals, ALARA cost benefit analysis and managen.ent involvenent.
a.
Historical Dose Goals i
For the years 1984 thru 1990, the licensee's annual Unit 1 and Unit 2 combined Dose Goals are presented in Table 7.
b.
Effectiveness in Tracking and Meeting Goals The inspectors noted that the licensee used pay for performance to stimulate ALARA awareness.
As one of the performance evaluation criteria a manager was evaluated on the achievement of the different goals using a sliding scale pay incentive.
In 1991, first line and mid-level managers will becomo more directly accountable for their accumulated doses.
Tables 8 and 9 describe the 1991 monthly collectise dose goals and department collective dose goals.
Table 10 is a representation of proposed collective dose goal broken down into monthly and outage /non-outage days and projected dose.
Accumulated collectivt dose and goal performance were reviewed during daily morning management meetings. Periodic memos from Plant Management to Departr;ent Managers help them evaluate the status of their Department collective dose goals.
The Health Physics and Chemistry Department provided the logistical support for record keeping and monitoring results, c.
Manageaent Involvement The unanticipated dose rate levels (three to five tiires expected) in the Unit 1 Dry Well resulted in a substantial overrun of the collective dose goals estimated for the recent outage.
Recognition that a Unit I source term reduction would have a significant impact on outage collective dose reductions, the licensee performed a cost benefit analysis to evaluate the chemical decontamination of Unit I recirculation piping / components during the next Unit 1 refueling outage.
The initial collective dose estimate for the Unit 1 outage was 555 person-rem.
Chemical decontamination should cause an estimated 366 person-rem collective dose reduction.
The project has been judged cost beneficial and monies for the Unit 1 chemical decontamination have been budgeted and a project team formed to undertake the project.
Administrative Procedure No 60AC-HPX-009-OS titled "ALARA Program" detailed the ALARA responsibilities of the plant personnel organization from General Manager / Assistant General Manager to plant
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The plant General Manager participated with higher level corporate management to establish a plant ALARA Goal.
Management has taken an active role in the Personnel Contamination Reduction Program.
From a level of 1867 documented personnel contamination events (PCEs) in 1986, the licensee has reduced the PCEs to 166 in 1989 and 183 as of December 8, 1990 (see Table 11).
The licensee credits this significant improvement to a seven point PCR Reduction Implementation Plan.
The Plan included the following key activities:
1)
Radiation Work Practice Guideline 2)
PCR Critique / Root Cause Determination 3)
Protective Clothing Dress / Undress Video 4)
Plant ALARA Review Committee Review of Contamination Trends
Health Physics Staff Reviews Trends with Contractors
Daily Management Review of PCR Status
Annual Performance Appraised tied to PCR's This reduction in PCEs from 1989 to 1990 was noted as a program strength.
The inspectors concluded that managenent involvement in setting annual collective dose 90als and management support of most ALARA initiatives was acceptable.
8.
ALARA/RWP Procedure laplementation Section 3 of RG 8.8 specifies that an effective radiation protection program to maintain occupational radiation exposures ALARA should include procedures, job planning, record keeping, special equipment and an operating philosophy.
The licensee used a radiation work permit (RWP) system to evaluate the radiological conditions and to specify the radiological control requirements to be implemented for radiological work.
Administrative Procedure No. 60AC-HPX-004-05, Rev.
6, dated September 25, 1989,
" Radiation and Contamination Control," defined the purpose of RWPs and established criteria for RWP preparation and approval.
There were two types of RWPs.
A " general" RWP was used for routine repetitive access to work in radiologically controlled areas (RCAs); and a " specific" RWP was required for specific jobs and where significant dose, contamination, or airborne activity may be involved.
Specific RWPs were valid for the duration of the job and it required by tne RWP, periodically reviewed during the job.
The inspectors reviewed the mechanics of ALARA reviews, pre and post job review criteria to determined adequacy of applicable ALARA controls and RWP reo,uirements.
The inspectors discussed with the licensee the methods used to relay ALARA controls and RWP requirements to workers and how the job dose was tracked.
The inspector noted that the licensee used input from job superviscrs and historical data to review ttam size and estimate the number of man hours per job, J
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The licensee's policies, goals and standards to reduce personnel radiation doses were specified in the "ALARA Program", Administrative Procedure No.
60AC-HPX-009-OS, Revision 5, dated October 2,1990.
The procedure also outlined the action levels for an ALARA review.
A review of the action levels indicated that they were effective in requiring attention to job with significant exposure potential.
HP technicians functioned as field ALARA coordinators during outages to review RWPs and serve as a point of contact between the workers and the ALARA staff.
9.
Plannirs and Scheduling Section 3.a. of _RG 8.8 specifies several considerations to be included in preparation and planning for jobs to ensure that exposures are ALARA.
'The inspectors reviewed outage reports over a five year period, interviewed managers and supervisors, and discussed outage planning and
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performance with licensee representatives.
Based on information from these sources, the inspectors noted that several problems in the planning and scheduling of outages required managetent attention, as discussed
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below.
Numerous sources indicated that increased work scope during outages was a I
chronic nroblem.
Table'12 confirmed the estimate of a 50 percent increase in work scope identified by one manager during interviews.
Prior to the 1990 Unit 1 Refueling Outage, the licensee established a scope control process.
After the initial outage work scope was determined, about three months prior to the outage, scope controls went l
L into effect for adding or deleting v.ork.
Maintenance work orders (MW0s) could be submitted by individuals.
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MW0s were reviewed by Planning and Controls to determine which-could be worked during outage versus non-outage scope.
The basic criteria was that
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when a component leaked air or water the. component would be fixed, Outage
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MW0s were either logged into the outage scope data bases by the planning l
and controls superintendent or. rejected based on an assessment of outage impact.
Logged entries received a second review by the outage manager to evaluate the impact on man power and the schedule.
An item would be added
.to the outage-scope and scheduled only after acceptance. All items added.
- rejected,- or deleted were reported _ to plant management and= various l
departments on a weekly MWO scope change report.
Feedback was evaluated by the Manager of.0utege and Planning for resolution.
For the Unit I refueling cutage, the plant received 93 design change requests (DCRs) for plant modifications.
The plant goal was to have completed all 0CRs six months prior to the outage and have all DCRs onsite i
three months prior to the outage start date.
During their review, the L
inspectors noted problems that may have contributed to the increase in outage work scope and high collective dose.
Licensee representatives stated that some DCRs arrived after the start of the outage and some did not have a previous ALARA review. The inspector interviewed personnel to
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determine the contribution of late arriving DCRs to collective dose or growth in outage scope, tut none of the groups had any quantified data.
The inspector also determined that there was no formal procedure foi including the ALARA staff in the review process during the developn.ent of
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DCRs. Licensee representatives steted that for the next outage there were plans were to' work 60 DCRs of which eight were anticipated to arrive late.
Based on discussion with licensee representatives, tinspectors determined that other problems existed that contributed to high collective dose and work coordination problems that the olent had not quantified or resolved satisfactorily.
Some of these problems included:
(1) overcrowding fn the drywell during the outage; (2) inclusion of HP earlier in outage planning; and (3) high amount of respirator usage.
Since radiological control problems were not identified in post outage reports, it was difficult for the inspectors to assess the radiological impact of these problems.
The inspector reviewed data maintained by the HP ALARA group for major tasks performed during the following outages:
Unit 1 - 1985, Unit 2 -
1986, Unit 1 - 1987, Unit 2 - 1988, Unit 1 - 1988 Unit 2 - 1989, and Unit 1 - 1990.
These date are summarized in Table 13.
The tasks evaluated by the inspectors were selected because the data were documented most consistently and were ccnsidered to be a representative sample of licensee performance in this area.
Previous problems in estimating dose for tasks had.been reported in audits and post cutage reports teith the estimation of person-hours identified as the primary problem.
Licensee representatives stated that efforts were underway to improve both dose and
person-hour estimates by -using historical data end a correction factor i
that would better represent the actual tine spent in the radiation area.
j 10. ALARA Initiatives / Operational Pract'ces j
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Hydrogen Water Chemistry (83728, 84750)
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During the outage for Unit 1 in the Spring of 1990, higher than
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Hydrogen Water
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Chemistry (HWC) _was suspected as being the leading contributor to
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these increased dose rates.
The current inspection partially focused on the progress the licensee had made in determining the causes of the high. dose rates: and also determining the ramifications of-the.
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high dose rates on the ALARA program at Hatch.-
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i As discussed in Inspection Peports 50-321 and 50-366/90-21 and 90-08, o
HWC had been implemented in Unit 1.
HWC control was achieved by the i
k injection of gaseous hydrogen in the reactor feedwater system.
HWC H
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is used to reduce the concentration of oxidizing species such as -
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dissolved oxygen (0 ) and hydrogen peroxide (H 0,) which are created L
by the radiolytic ddcompositico of water in th8 t'eactor core.
These l.
species are considered to be contributors to intergranuhr stress
corrosion cracking (IGSCC) in the reactor coolant system (RCS) piping and welds.
The hydrogen combines with the oxygen in the downcomer l
region of the reactor vessel, causing a decrease in - the
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electrochemical potential (ECP) of the stainless steel piping, and o
reducing the potential for IGSCC (oxygen is a major contributor to
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IGSCC).
ml During power operations, nitrogen-16 is continuously being formed in
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the reactor core from the neutron activation of oxygen-16.
During normal water chemistry (NWC) in a BWR (i.e. no hydrogen injected)
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most nitrogen is in the form of nitrates or nitrites which are
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non-volatile and tend to remain in the bulk reactor water.
Some carryover does occur which results in normal main steam line (MSL)
e radiation levels.
During hydrogen injection a reducing environment F
is created in the bulk water.
Nitrogen is converted to anmonia, p
which is volatile and is carried over to the MSLs resultirg in signif' cant increases in MSL radiation levels.
If these increased r
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r% tion levels are too high, the hydrogen injection rates may be
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limited.
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f The licensee provided the inspectors with information from their contractor (General Electric) that indicated that gan:ma spectroscopy
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data supported the theory that the increases in radiation levels may
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be caused by releases of particulates from the core. Water chemistry
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changes may increase the releases of activated species from the core.
These releases may have been from fuel deposit origin, or from a
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corrosion film release.
These species could be redepositing on pipe
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Surfaces, contributing to the increased dose rates. These releases
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were thought to be relatively short-lived (less than six nonths),
lasting until the RCS reached equilibrium with the changed chemistry
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environment invokec by the use of the HWC.
Studies have indicated that dose rates under HWC increase initially, and then appear to
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reach a plateau General Clectric theorized that cycling HWC on and
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off may extend the length of tin,e and extent of these releases, and further increase dose rates.
Zinc irjection may not mitigate this problem significantly, since zinc injection only affects the
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re-distribution associated with the use of HWC.
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When HWJ is implemented, the RCS changes from an oxidizing to a reducing environment.
This shift in chemistry effects the oxide layers in the systen, and the cobalt-60 buildup rate may increase
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until the oxide layer comes in equilibrium with the system.
Under f,
normal water chemistry, in plants with primarily 304 stainless steel
materials, hematite is formed.
Hematite forms a thick oxide layer, y-and has few natural crystal sites for cobalt ions.
In a retural zinc l
plant, zinc might preferentially fill these sites, aiding in keeping the radiation fields at a low level. Under HWC, it is theorized that i
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the oxide layer may shift to spinel. Cobalt is readily incorporated Q
into Spinel.
i General Electric (GE) has recommended that, to reduce effectively the
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potet,tial of IGSCC, an ECP of -230 rnillivolts (mV) Standard Hydrogen
Electrode (SHE) snould be maintained between the bulk water in the l
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RCS and the pipe walls. Previous testing at Hatch has shorn that a hydrogen injection rate of 45 standard cubic feet per minute (SCFM)
was required to maintain an ECP of -230 n.V SHE.
However, at this injection rate, the radiation levels in the MSts increased significantly and caused the general dose rates arcund the plant to be excessive.
Because of this, the hydrogen injection rate was reduced to 22 SCFM.
An injection rate of 22 SCFM resulted in an ECP of -100 to O my SHE, greater than what was recomended for optimum protection against IGSCC, but vill less than the ECP obtained without hydrogen injection (+180 mV).
Testing showed that the increased dose rates at 22 SCFM were tolc "ble.
During the Unit 1 aring 1990 outage, the licensee replaced the admiralty brass na i condens" tubes with titanium tubes.
This replacement was inte ded to reduce incidences of condenser in-leakage and copper induced localized cracHng (CILC) of the fuel cladding brass contains large amounts of copper.
Copper, in since admiralty (Cu4+), can act as an oxidizing agent. similar to its ionic form
and 0 as discussed above.
Erosion vi the admiralty brass conden
tubes resulted in measureable levels of copper in the RCS. This contributed to the higher hydrogen itjection rates which were required to maintain the recomended ECP.
During the ongoing cycle, following the outage, hydrogen injection rates were reduced to 16 SCFM and an ECP of -450 mV SHE was obtained, due to the reduction in ionic copper in the RCS.
Tra MSL radiation levels were reduced from 2.0 times background w;th an injection rate of 22 SCFM, to 1.8 times
background with an injection rate of 16 SCFM.
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Admiralty brass also contains trace amounts of zinc, and the erosion of the condenser tubes resulted in reactor water z int concentrations of approximately 10 parts per bi'. lion (ppb) with hydrogen injection, and 20 ppb without hydrogen injection. Previous industrial research and experience has indicated that maintaining ionic zinc at these concentrations in the RCS should significantly reduce activated corrosion product deposition, especially cobalt-60, on primary system piping walls.
HWC was supposed to increase the efficacy of this effect.
This apparently was not the case at Hatch.
Gamma spectroscopic studies of the Unit I recirculation piping oxide coating during the Spring 1990 outage indicated a cobalt-60/ zinc-65 activity ratio to be approximately 3 to 1.
Previous similar, informal studies for Unit 2, which had not been under HWC control indicated a ratio of 1 to 1.
The cause of this increased plateout was still under investigation by the licensee and by GE at the time of this insnection.
The reactor water zinc concentrations during cycles 11 and 12 may have been 'nsufficient to inhibit colbalt-60 in the oxide film.
The licensee impleaented zin. injection in both units in August 1990, to help control radiation field buildup in the recirculation piping.
Prior to the Spring 1990 catage, Unit I had zinc in the RCS due to the erosion of the admiralty brass condenser tubes (see above). Unit
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2 had.nevar been ur e HWC, and had the concenser tut >es replaced with titaniba ir the Fen of 1989.
The effects of the zirs injection shtuld betonie cvident during the next outages for these units; in March 1991 for Unit 2, and in Septen:tser 1991 for Unit 1.
The exact cause(s) of the increased Unit I drywell dose rates has not yet been deterrrined.
One reason for this is that the chemistry of the system hes been changed nun +rcas times, naking it extrenely difficult to determine which changes to the systen caused the increase.
For several cycles Unit 3 was run under tiWC, with admiralty brass condenser tubes, causing it to be a "r.atural" zinc plant.
Then, during Cycle 10
">it I underwent HWC testing, and was cycled back and forth under
"'t rad HWC at varying injection rates and times; still a ratural 21.x Nnt.
HWC was implemented during the latter part of Cycle 11.
n.ecirculation piping cose rate had shown some increase by the enc of Cycle 11.
Unit I was brought online for Cycle 12 under NWC for a short period of time (apyoximately 50 days) before being switched to HWC for
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approximately 374 days.
During this portion of the cycle, HWC was cycled between 12 and 22 standard cubic feet per minute (SCFM)
approximately three times per week in order tc reduce the nitrogen-16 concentrations during maintenance and surveillance activities in the
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condenser t'ay.
There were some occasions when there was a zero SCFM injection rate. Unit I was still a natural rinc plant for this cycle.
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The Spring 1990 outage followed Cycle 12 which was one of the
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longest sustained power runs rece*ded for a plant under DC. It was during this outage that the high drywell dose rates were discovered (higher than that following Cycle 11).
During this outage the condenser tut'es were replaced in Unit I with titanium tubes.
Again, the unit was brought back online for Cycle 13 under fiWC, this time with only residual zinc in the pre-existing oxide layer.
Zinc injection was started in August 1990, and a short time later the plant was put under HWC, As before, the HWC was cycled between 12 and 16 SCFM for the reasons mentioned at,ove, with short periods of time at zero SCFM.
Mid-Cycle 13 measurements indicate that there have been no further increases in dose rates.
At the time of the inspection, Unit I was back under NWC, with zinc injection. With all the changes being made to the system, it would be very difficult to determine what caused the high dose rates in the drywell, in addition, there is o need for information on the morphological changes of the oxide layer as a result of the changing chemical environment.
Ganna spectroscopy was performed to obtain isotopic analysis on fuel scrapings taken during outages; and on pipe surfaces
during outages and at other times.
This provided information on the
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relative amounts of radioisotopes present, which was very valuable, l
but only part of the equation.
These analyses would not indicate the l
chemical form, or oxide matrix, in which these species existed, or tell much about the changes to the oxide layer as a whole.
The licensee was considering taking oxide layer samples off of a flange located in the recirculation piping, to attenpt to obtain this type
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of data.
This would be performed for Unit I during the September, 1991 outage, and for Unit 2 during the March,1991 outage.
Since Unit 2 will not have been under HWC, these samples should provide good baseline data.
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Discussions with and documentation provided by the licensee indicated
that at least five other plants which implemented HWC had shown some increase in dose rates, including hotspots.
These plants were Brunswick, Dresden, Duane Arnold. Fitzpatrick, and Morticello.
Brunswick's Unit 1. recirc piping dose rates increased by a factor of three to five over normal levels after four months of HWC.
Brunswick's Unit 2 recirculation system everage general area dose rates in milli-Roentgen per hour (mR/h)(decon), and pre HWC; increased from approximately 85 mR/h, pre-chemical decontamination to approximately 350 mR/h, post HWC.
A hard scram at full power, and the --loss of recirculation flow, may have cautad a crud burst and allowed an un.
.11y high amount of plate-out of insoluble cobalt.
Monticello found recirculation pipe dose rates of approximately three times normal after running under HWC (no zinc injection) for seven months.
Monticello perfctmed a chemical decontamination during a refueling outage.
The present cycle was begun with HWC and zinc injection.
At eleven months into the current cycle, the dose rates were approximately half of what was seen with HWC without zine injection.
Fitzpatrick performed a chemical decontamination during refueling and prior to inplementing HWC and zinc injection.
During the following refueling, dose rates were approximately three times expected levels.
Plants thich had natural zinc components in their reactor water due to the materials used in construction of the plant, and plants which had implemented zinc injection, showed the smallest increases.
Hatch performed a chemical decontamination of the reactor water cleanup system (RWCU) during the Spring 1990 outage. The dose rates around the.RWCU were reduced five to ten fold, The licensee has conducted dose rate surveys of this area during the current fuel cycle. These surveys indicate, that while dose rates have increased, they have not. increased to pre-decontamination levels. Citric acid and potassium permanganate were used during the decon process.
The licensee plans on performing a chemical decontamination of Unit I recirculation piping during the next outage, scheduled for September, 1991.
This decontamination should significantly reduce dose rates and occupational radiation exposure during the outage.
The licensee plans on using 'a low-oxidation state metal-ion process to accomplish the decontamination.
The inspectors attended one of the initial meetings held by the licensee to plan and organize this process. The
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licensee planned on having the vendor use metal coupons, representative of the materials in the system, to track the effects
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of the decontamination process on system materials.
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The licensee had originally intended to implenent HWC control for Unit 2 in December,1990.
A refueling outage was scheduled for
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March, 1991.
The licensee decided to deloy the implementation of the i
HWC until the next cycle, becase there was not enough time to organize and schedule a chemical decontamination of pertinent sections of tha RCS priar to the scheduled outage.
The licensee recognized that performing a chemical decontamination con eliminate a significant amount of occupational exposure during the outage.
Hetch has used Crack Arrest Verification Testing (CAV) to determine the effectiveness of the HWC.
The ECP and CAV measurenients were being performed in autoclaves external to the RCS, in a simulated environment.
There has been son;e question as to whether these test
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results were trt.ly representative of material conditions in the reactor environnent. Although current test results of precracked test s.,ecimens in the CAV showed no significant crack growth, evidence of pipe cracking has_been discovered by the licensee, it was suggested by GE that the licensee's techniques for identifying cracks has improved.
Since the reactor core is a more severe environment than the recirculation piping, it has also been recognized that an ECP of-230 mV SHE in the recirculation piping does not guarantee that the reactor internals will be protected against IGSCC.
GE has made several reconnendations to Hatch concerning the imrlementation and operation of HWL cort.'ol. Sone of these
reconcendations included:
Rur, the units under continuous HWC, avoiding cycling of the
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hydrogen injection rates.
Maintain the reconeended levels of zinc in the reactor water.
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Develop plans for chemical decontsmination for future outages.
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Consider the installation of in-pipe ECp probes in the
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recirculation system and bottom drain line to confirm CAV'and ECP measurements.
Install in-core stress corrosion cracking monitor to determine
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reactor internals HWC prot?ctior rates.
Monitor and trend soluble and insoluble radioisotopes in reactor
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water.
On January 24, 1991, a meeting was held at the NRC's HQ office with licensee representatives to discuss the status of the HWC program at
. Hatch.
The information presented at the meeting included Hatch's chemistry control, operating issues of HWC, Hatch's experience with HWC, and future plans for HWC at Hatch.
In attendance were personnel from Hatch, Georgia Power Corporation, and General Electric; as well as regional inspectors, and Headquarters personnel, i
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NRC representatives discussed with Hatch personnel the current status of the HWC program.
HWC control had been temporarily halted at Hatch on Unit 1 in mid-November, 1990.
This was due t: 'ow oxygen readings from the Offgas monitor and high MSL radiation rates, causing a near trip on the reactor. The licensee stopped injecting hydrogen and
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initiated an investigation to determine the cause of these events.
It was originally believed that a mechanical failure of the hydrogen injection system allowed a greater than expected hydrogen injection rate, and that the excess hydrogen con.bine with the oxygen, lowering the oxygen readings from the Offgas monitor.
The investigation did not find any evidence that this was the case.
It is now believed that small amounts of loosely held organic
material was building up on the resin beds in the RWCU System.
Changing out a resin bed would temporarily cause increased flow through the remaining beds, causing the organics to " wash-off" the
,
resin beds.
These organics, suspected to be short-chained hydrocarbons, would break down in the reactor core, producing hydrogen.
The licensee was still involved in determining the source and nature of the hydrocarbons as of January 30, 1991.
Discussions with the licensee indicated that Hatch was planning to
,
restart HWC control on Unit 1 in early February,1991.
The plans were to increase hydrogen injection rates to 35 SCFM 10 provide protection for lower reactor internals.
The amount of hydrogen injected 'was determined by the amounts needed at Fitzpatrick, a sister plant to Hatch which has an incore monitor; and by calculation performed by GE using peroxide concentrations in pertinent portions of the system.
The inspectors, as best could be determined within the scope of this review, considered that adequate forethought and planning went into the myriad decisions that confronted the licensee concerning this complicated issue; and that the licensee bad made the best of a
" learning" situation.
The licensee werked closely with their contractor during this process, and used-their contracto'r as a resource to identify and resolve problems and issues as they arose, in addition, based on this review, the inspectors centidered that the HWC control program was being operated in a manner co3 current with ALARA principles.
-b.
Source Tenn Reduction Program (83728)
The inspectors also reviewed the actions the licensee had taken to ensure fuel integrity and to reduce source term.- Some of the areas covered under these efforts included:
reduction of the transport _of copper and and other metals; reduction of air inleakage and radweste inleakage; feedwater oxygen optimization; adoption of EPRI "BWR Water Chemistry Guidelines;" condensate tube replacement; layup of major plant equipmeit and zinc injection.
Several of these areas have already been adt.9ssed in previous sections of this report (HWC,
,
D
.
,u,..
....,
-
.
.
,
,
- zinc injection, condensate tube replacement, and decon of the retirc system).
The inspectors reviewed graphs covering the time frame of mid-1989 to September, generated by the licensee, that illustrated a decrease in insoluble copper in the feedwiter for Unit 1. The concentration dropped from approximately 0.03 ppb copper to approximately 0.01 ppb copper.
Feedwater soluble copper was also reduced for this tine freme, from approximately 0.4 ppb to 0.1 ppb.
The inspectors reviewed graphs which illustrated, that for Unit 1, cycles 8 through 11, that feedwater iron had decreased from approximately 11 kilograms iron input per one hundred thousand megawatt-days to approximately 5 kilograms iron input per one hundred thousand n,egaWatt-days. For Cycle 12, the feedwater iron was approximately 10 kilograms iron input per one hundred thousand megawatt-days.
Unit 2, currently in Cycle 9, showed a slight decrease f or this p6raneter for fuel cycles 4 through 7, and then a slight increase in cycle 8, to approximately 6 kilegrams iron input per one hundred thousand negaWatt-days.
The inspectors also reviewed graphs depicting liquid inleakage for equipment and floor drains for 1988 through December 3,1990.
The graphs show, that for both units, that liquid inleakage was reduced from approximately 60,000 gallons per day in 1988 to less than 20,000 gallons per day in 1990.
Air inleakage was also reduced for both units for this time franie from approximately)40 SCFMless than 20 SCFM per day (pre per day to for Unit I had also been reduced, from approximately ten ppb in 1989, to approximately two ppb in 1990.
The inspectors also reviewed chemistry information that showed that the time the plant was out of EpRI "BWR Water Chemistry Guidelines" decreased frcm approximately 14 percent in July of 1988, to 0 percent for Marc l through November, 1; 1 (graphed on a monthly basis).
The inspectors also determined that the licensee had a procedure that provided instructions for the dry layup of the condensate and feedwater systems for Unit 2.
The inspectors determined that the licensee also had made some material replacenents with low cobalt containing alloys. The inspectors reviewed Health Physics / Chemistry Departmental Goals for 1990 and determined that several of these goals incorporated the source reduction actions discussed above.
Although the licensee had taken several steps to reduce source term and ensure fuel integrity, there was not a formal program which encompassed all of these actions, listed long term goals, or tracked actual source tenn reduction in terns of person-rem saved.
The inspectors did review a " Top Ten Problems at Hatch" list. While this list did include items which would result in source term reduction, it was not deteiled or extensive enough to be considered a source term reduction program, it should be noted, however, that by
.
--_---___._.i_._________________m-_.-_
.
'
'
.
deconning the recirculation piping prior to outage activities, that significant source term reduction should be achieved.
11. Assessment /Self Evaluations a.
Audits Section C.1.b of RG 8.10 specifies that licensee managenent should periodically perform a formal audit to determine how exposures might be lowered.
The inspectors reviewed ALARA related audit reports, including Audit Report No. 88-HP-2. The audit reports supplied to the inspectors at the beginning of the week by the licensee did not contain any responses to the audits.
The inspectors were able to obtain responses to the audits from the licensee but some were supplied after the exit interview and a review of these audits and responses was not made until the following week.
The following is a sumary of findings identified during the 1988 88-HP-2 audit of the ALARA program:
1.
Evidence could not be found that PARC is an active / effective committee.
2.
Workers do not know that ALARA is everyone's responsibility.
3.
Some supervisors do not know some ALARA techniques.
4.
None of the workers interviewed knew their departmental ALARA goals.
5.
The initial man-rem estimates for RWPs are inaccurate in that actual dose comenly totaled two to three times greater than the estimated dose.
6.
Problems causing more exposure to workers than necessary were:
inadequate surveys, inadequate planning, inaccurate HP coverage requirements on RWPs, and lack of communication between HP technicians and workers.
The audits were conducted by the licei ee's Quclity Assurance Auditors who were qualified to perform audits in these areas.
The inspectors determined from a review cf these areas that the licensee's corrective cctions instituted for these findings were not effective in correcting the problem in all areas. Items 1,2,4 and 5 above were noted again during the current NRC inspection as areas needing improvement; therefore, the items were not adequately corrected as stated in the " Audit Finding Closure Letter" dated April 12,1989.
The inspectors concluded that the licensee had corrected the problems identified in 88-HP-2 within a limited scope and had not determined or corrected the root cause of the problem.
.
_ _ - _ - _ _. - - - - - - - _ _ _ _ _. _ _ -. - _ -
_-
i
'
.
t
.
'
The inspectors notified the licensee that improvements to corrective actions to CA Audit findings would be identified as an improvenent area.
Responses to QA audits will be reviewed during future NRC inspections, b.
Post Job Reviews Post-job reviews were conducted after each job for which there was an ALARA Review Package completed.
The review included a comparison between the budgeted and actual man-rem and man-hours for the iob und a brief analysis-of the job frem the ALARA standpoint.
Post-job review meetings were held for all jobs which had exceeded five man-rem.
Workers, supervision, pre-job planners, and other eersonnel involved usually were required to attend.
Critiques of the job and discussions for future dose reduction techniques were discussed.
Records of the meetings were kept on the ALARA Post-Job Review Sheet.
The inspector concluded that the meetings were substantive and provided information that could be used during future similar jobs.
c.
Maintenance Reworks The inspectors reviewed selected maintenance records for rework activities that resulted in unnecessary collective dose.
The inspectors determined the licensee identified root cause for rework activity.
Records indicated that in 1986 nine percent of all maintenance work was rework.
This number improved to 1.0 percent in 1989 and was currently 1.2 percent for 1990, in Novenber 1990, five percent or eight welds were rejected.
Two rejections were for paperwork problems and the other six rejections were due to bad welds.
The licensee determined that all were performed by the same welder.
This welder was currently being retrained and the licensee was holding " tool box" sessions with the other welders.
12.
ConfirmatoryMeasurements(84750)
As part of the NRC Confirmatory Measurements Program, spiked liquid samples were sent on January 6,1989, and on October 8,1990 to the plant for. selected radiochemical analyses.
The NRC received the analytical results -from Georgia Power Company in letters dated March 3,1989, and November 29, 1990. The comparison of licensee results to known values are presented in - Table.1d.
The acceptance criteria for the comparisons are listed in Table 15.
The results for 1989 were in agreement. The results for 1990 were not in agreement for the strontium-89 and 90. The isotopes in disagreement were reanalyzed and the results reported to the NRC via telephone on January 2, 1991.
The result for strontium-89 were in agreement, while the result for strontium-90 remained in disagreemen __
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.
.
.
.,
- The cause of the disagreements was not known at this time. This area will be reviewed during a subsequent inspection.
Additionally, an iron-55 sample was provided to the Georgia Power environn1 ental lab located in Smyrna Georgia.
This lab planned on performing iron-55 analyses in the future.
The result for this analysis was reported to the NRC in a letter dated November 12, 1990, and is presented in Table 14 The acceptance criteria for the comparison are listed in Table 15. The result was not in agreement.
The cause of the disagreement was not known at this tine.
This area will be reviewed during a subsequent inspection, 13.
Exit Meeting The inspectors met with licensee represcritatives (denoted in Paragraph 1)
at the conclusion of the inspection on December 14, 1990. The inspector sunnarized the scope and findings of the inspection, including the ALARA program improvement items.
The inspectors also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspector during the inspection.
The licensee did not identify any such documents or processes as proprietary.
Dissenting comnients were not received from the licensee.
Acronyms and initialisms ALARA as low as reasonably achievable CAV crack arrest verification CRD control rod drive DCR design change request decon decontamination D/W drywell ECP electrochemical potential GE General Electric GET general employee training HWC hydrogen water chemistry Hx heat exchanger
>
IGSCC intergranular stress corrosion cracking ISI inservice inspecticn LPRM low-power range n.onitor mR/h milliroentgen per hour MCL main steam line Mf materials test mV millivolt VW0 maintenance work order NDT non-destructive test NWC normal water chemistry 0JT on-the-job training PARC Plant ALARA Review Committee ppb parts per billion PT penetration testing
_ _ _____ ___ ____________________--_______________
_
_ _ _ _ _
-_-
_ _ - _ - _ _
'
.
.
,
,
'
R C,'
reactor coolant system RG Regulatory Guide RWCJ reactor water cleanup R,
reactor SHE standard hydrogen electrode SRM source range monitor CT ultrasonic testing
.
,
?
-_.. _ _ _ _ - _ _ _ _ _ _ _. _.. _. -. - _ _. _ _ _ _ _ _ _. _ _. _ _ _ _ _ _ _ _. _ -. _ _ _ _ _ _ _ _ _ _ _ _ -. - _ - _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ - _ _ - - - _ -. - _
___
... _ _ _
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.. _.. _ _. _.. _ _ _ _. _ _. _ _ _ _ _
-. _ _.. _ _ _. _ _ _. _.. _. _.
-
.
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=
.
.-
,
,
.
.
,
TABLE 1
'
Lea,r,1y,CollectiveDoseLUnitforU.S.BWRsCompary,dwithHatch r
Year Person-Rem /j{n,i,t (1)
Hatch / Unit Average 1984 1003 1056 1985-735-409 1986
'651 835 1987 527 408'
1988 511 700
.1989 442 334 750(2)
1990
Notest Not yet available
NUREG 0713 Vol. 9. " Occupational Radiation Exposure at Connerical Number Power Plants and Other Facilities 1987 Table 4.1
As of December 8,.1990
'i
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.. - -..
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i l
TABLE 2 Actual versus Estimated Collective Dose for Major Jobs i
Unit 1_ Outeg,e, (9/29/8_4_ - 1/_13/65]
Actual Estimated Job Description /
Dose Dose Actual Estimated Location
_(person-rem)
(person-rem)
Hours'
Hours Decon D/W 34.9 4.0 2319
Insulation D/W 28.5 10.0 1007 200 HP Coverage D/W 11.0 32.0 853 3200 Weld Testing 58.7 3.0
- 2563
UT, MT,-PT, NDT D/W Install Insits.
10.9 5.0 717 100 llP Coverage R/X 14.2 10.0.
5220 5580 Scafford/ Shield D/W
?3.5 15.0 829 600 Snubbers D/W 28.5 5.0 2273 560 Inspections 0/W Eng.
25.8 0.1 1057 100 Snubbers / Toras 17.3 0.5 4751
Repair Valves D/W 10.4 7.6 1420 1540 Insta11' Conduit D/W 33.2 1.0 1479
Welding On Recirc D/W 78.0 24.0 29',2 800 HP Coverage 10.9 9.0 2922
. 5580 HP Drywell 10.2 32.0 1199 3200
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v--c-rv--mm,-
ow-e+saw-g'W-ar=Marewwe*W-'==
- w su--r
-*
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-
.
-
_ _ - - - -... -. -..-.-..-..-._-
-- -.-..
.
_
.
'
'
.
,
TABLE 3 Actual versus Estimated Collective Dose for Major Jobs UnitIL2,0utaje_(,1,1L2.7/85-5/3/861 Unit 1 Job Description Actual Estimated Actual Estimated Location Dose (per,s_o,n, rem)
Dose (person-reml Hours _
Hours
_
-ISI/DW Recire 20.6 22.0 906 950 Piping IHSI D/W 15.0 13.5
450
>
Inspection D/W 10.5 10.3 643 650 HP. Coverage D/W 23.8 16.0 1482 1600 Scaffold D/W-11.2-12.0 930 950 CRD /DW Rem & Trans 21.2 10.0 956 600 We H DW Overlays 22.8 21.0 873 780 Insulation /DW R&R 31.8 48.0 1431 2000 Buffwelds D/W ISI 21.6 24.0 721 800 Valve Repair D/W 10.1 12.5 1997 2500 R&R Install RX 11.1 12.0 3102 3000 Electrical Comp.
. Install Piping D/W 29.5 35.0 2422 9292-
-
& Hangers IHS1 On D/W E11 95.900 100.0 3840 400
& B31 Unit 2 Job Descriptio'n Actual Estimated Acual Estimated
.
Location Dose (person-rem]
Dose (person-rem)
Hourr, Hours
-
HP Coverage D/W l'. 2 11.8 2728 2430 Decon D/W 12.1 11.6 1593 2400
' L HP Coverage-D/W 19.3 21.1-3275 3325 Insulation D/W.
22.2 21.6 2107 1950 1HS1 D/W 11.2 12.0 614
.1500 LPRM, SRM, IRM 17.6 22.9'
492 635'
CRDs 11.4 11.4 70; 725-l l-
.
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_ _ _ - _ _ _ _ _ - - _ _ - _. _ _ _ - - _ - _ - _ _ _ _ _ _ _.
. _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _
.
.
,
.,
TABLE 4
Actual versus-Estimated Collective Dose for Major Job; Unit 1/2Outa_ge{1988]
Unit 1
'
Job Description Actual Estimated Actual Estimated-Location-Do,sL(person-rem)
Dpo_sp,p vson-rem)
Hours Hours Decon D/W 21.5 13.0 1650 1300 Shielding D/W-20.8_
7.2.
688 350 Remove /Rapiace 14.7 5.0 327 100 IRM/SRM/LPRAS CRDs
.
10.9 15.0 609 500 HP RX BUILD-13.9.
12.0 2524 2000 Routine Miwor R/X 11.8 9.0 4012 3000 MOD Supports D/W.
39.4 1.0 2611 500 insulation D/W 62.4 20.0 2630 2000 IS BB1 SYS D/W 22.8 8.0-850 400 Painting D/W-42.2 4.5 2621 100 Elect CAL D/W 16-'3 4.5:
781 300
-
.
Remove / Replace-21.0 0.6 728 300 631 Pipe RWCV HX
.
-Remove / Replace 63)
55.9-15.0 1254 800
'
Drywell -
Replace 1631-Coola 31.9 12.0 1554 1000
. Pump Rm "A" HP-Coverage R/X.
12.8 25.0 5853 5000 Pipe Support MOD:
14.0 0.5-3026 400 Torus Scaffolding D/W.
21.0 3.5 1273 600 Clean Cooling Coils 10.2 0.5'
428 140.
D/W Unit 2-
~
-Job Description LActual Estimated Actual.
Estimated
.Lo;ation Dose (person-rem)
Dose (person-rem)
Hpurs Hours on s
HP Coverage R/X
. ' 10t. 7
. 8.0:
3345 1000 Vessel Disassembly 14.3 12.0_
939 600 HP Coverage D/W 13.)
15.0 2436 300
- Decon D/W 15. 'c 4.0 2125 160
- Insulation D/W 29.5
'5.0 1953-300
'
CRDs.
12.7 13.5 789 300 Routine Minor D/W~
-12.2-20.0 1676 2500-Install Supports, 10.9 2.6 1487 350 Conduit'
.ISI Support Work 11.1 4.0 417 450 L
.,
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-.
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-
-
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TABLE 5
'
Actual versus Estimated Collective Dose for Major Jobs Unit 1 Ou,tage,,(19,9,0)
Job Description Actual Estimated Actual Estimated Location Dose (pe,rson,-rem)
Oose(perso_n-rp,m)
Hours Hours HP COV (Rx)
14.6 11.0 2428 3500 OPS lits (Rx)
10.8 9.6 1464 1600 Routine Mech (Rx)
13.7 14.0 3915 7000 631 Hx Repair *
41.3 75.0 934 1650 HPCOV(DW)
62.8 35.0 5310 4200 Insul. (DW)
65.1 71.0 2877 3000 Shielding)(DW)
48,1 21.0 1597 700 79-14 (DW 59.8 48.0 2821 4200 151 (DW)
63.2 33.0 2068 1200 (B31 MTR R/R(DW)** 62.3 65.0 4233 4020 CRD R/R (DW)
14.0 16.5 758 800 Elect. S. Pile (DW) 13.6 10.2 334 300 B31 Pump inp.(DW)** 13.4 18.0 825 1000 Minor Elec. (DW)
14.2 16.0 876 1110 Snubber R/R DW 16.6 4.8 623 280 Decon (DW)
43.4 50.0 3025 3500 Scaffold (0W)
50.9 14.0 2897 1300 79-14(RX)
22.4 18.0 4461 5000 B31 Overlay (DW)** 20.4 80.0 4971 3000 D/W Chiller RR(DW) 154.2 118.
7017 5120 R. Minor Mech (TB)
14.9 5.0 43485 14000 Retube Condenser 13.6 15.9 33370 35500
- 631 = Reactor Water Cleanup System
- 831 = Recirculation System l
-
-
_ -- - - - - - _ - - - - - - - - - - - - - _ - _ - _ _ - _
-- _
_._..-.. _. _..._ _.... _ _... _ _ _ _.. _ _. _ _ _. _.. _. _. _.. _ _ _ _.. _ __.. _ _._ _._.,
i
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.
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'
.
l TABl.E 6
$$IR9ff_IdSAtip!Lla,ttley,sy,s_Raig.S,tpgL}ingpadiationMonitor l
,
Hydrogen Injection Rate StandardCubicFee_t_perMinute,(SCFM)
Dose _RatL(Re_
>
1.3
1.3
2.2
4.1
-25 5.1
5.5
5.6
- 5. 8 42.
5.8
,
f
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. -. _,. _,...... _ _.,., _...
-.. _.. _
.
.
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..
.
.
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. _. = _
_
.. _ _ _. _. _
._. _._. _ _ _._.._._
_. -
_ _.. -. _ _.
.
.
.,
,.
.
TABLE 7 Performance Dose Goals Vs Actual Collective Dose
Outage Performance Actual Duration Station Goals Collective Dose Year (Days)
Unit Goal (person-rem)
.(person-rem)
1990 109 U-1 1234 Level 5-1050 1502 as of 12/10/90 Level 4-1155 Level 3-1270 Level 2-1525 Level 1->1525 1989 105 U-2 494 Level 5-900 668 Level 4-1050 i
Level 3-1200 j
level 2-1400
,
Level 1->1400 1988
U-1 762 Exce11ent-1060-
U-2 308 Commer.dable-1200 1401 Fuliy accept.-1500 1987
U-1-435 Excellent -1060 Commendable-1100 848 Fully accept -1500 1100 1423 1986
-
1985 167 U-1 1070 Excellent 1400
- 50 U-2 Commendable 1300 828 Fully accept.-1200 1984 106 U-1 Excellent 1980 235 U-2 Commendable 2200 2112 Fully accept,-2400
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,
.
TABLE 8 1991 Monthly Collective Dose Goals
. i. 5 LEVEL 4 LEVEL 3 LEVEL 2 LEVEL 1
,
MONTH rixs0N-REM PERSON-REM PERSON-REM PERSON-REM PERSON-REM JANUARY
30
33
>33 FEBRUARY
27
30
>30 MARCH 117 122 128 138
>138 APRIL 251 261 271 292
>291 MAY 228 238 248 267
>267 JUNE
29
3.'
>32 JULY
30
33
>33
AUGUST
30
33
>33 SEPTEMBER 103 107 111 119
>119 OCTOBER 207 216 224 242
>242 NOVEMBER 149 155 161 174-
>174 DECEMBER
30
33
>33
. TOTAL 1225 1275 1325 1425
>1425 I
l-
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.
.
.
.
_
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. _.
.
.
.
..
.
.,,
.,
TABLE 9 1991 Annual Department Collective Dose Goals LEVEL 5 LEVEL 4 LEVEL 3 LEVEL 2 LEVEL 1 DEPARTMENT PERSON-REM PERSON-REM PERSON-REM PERSON-REM PERSON-REM OPERATIONS 65.0 67.0 70.0 75.0
>75.0 MAINTENANCE 886.5 923.3 958.9 1031.0
>1031.0 HP/ CHEM 176.0 183.0 190.0 204.0
>204.0 ENGINEERING 69.0 72.0 75.0 81.0
>81.0 GEN. SUPPORT 18.5 19.1-20.0 22.0
>22.0 TRAINING 0.4 0.4 0.4 0.5
>0.5 QA 0.4 0.4 0.5 0.5
>0.5 SECURITY 5.5 5.8 6.0 6.5
>0.5 O. & P.
3.6 3.9 4.0 4.3
>4.3 NSAC 0.2 0.2 0.3 0.3
>0.3 TOTAL 1225.0 1275.0
'1325.0 1425.0
>1425.0
,
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.-
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.. +.., -
-
.. -
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
__
_ _ _ _ __
. _ _.
_
_ _ _ _ _.
_ _ _ _ _ _ _ _
. '.
.
.
-
.
'
,
TABLE 10 1991 Collective Dose Goal (Proposed)
Days Month Non-Outage Outage Collective Dose (person-rem)
January
0
February
0
March
12 128
.
April
30 271 May
27 248 June
0
July
0
.
August
0
,
September
13 111 October
31 224 November
21 161 December
0
Total 231 134 1325 1.
Non-Outage Goal:
1 Person-rem / day 2.
Unit Two Outage (03/20 - 05/27 est.):
Drywell Person-Rem: 400
.
Outage Total Rem: 624 1.
l 3.
Unit One Outage (09/18 - 11/21 est.):
l.
Drywell Person-Rem: 170 Outage Total Person-Rem: 470 l
i l
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.-..
.
-
_
__
.
.
. - - -
___ - ____ _-__
..
,
4
.
.
TABLE 11 Personnel Contamination Events (PCEs)
Year PCEs 1986 1867 1987 588 1988 1120 1989 166 1990 183
.
._
_ _ _ _ _ _ _. - _ _. _ _. -. _ _ _ _ _ -. _ - _. _ _ _ _ _ _ _ _ _ _ _ _ _ _.. _ _ _ _ _
__
_ _ _
_
._.
...
.
....
.
.
TABL'c 12 Refueling Outage Maintenance Work Order History Year / Unit Original $ cope Added Scope Work Hrs Total
% Scope 1986/2 1447 2277 3724 157
-
1987/1 1037 749 1786
-
1988/2 1405 404 429 2238
1988/1 1476 982 390 2848
1989/2 1888 816 375 3079
,
1990/1 2468 725 590 3783
l
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TABLE 13
,
e
,.
ALARA Estimates /Actuals for Selected Outage Tasks Task Dose Est.
Dose Lt.
Percent Man-hours Est,, Man-hours Acts Pe rcent DifTerence
{ summerized
[pe rson-remi ipS rson-reml D1fference over 7 ouja ge_s.)
HP Coverage 59.7 90.2
+33.8 13,510 17,7t:6
+23.9 ieeactor Building I t 7. 84
+ 9.1 18,325 18,097
- 1.26 HP Coverage 134.0
,
!
Dryve I I I
l Drywell Decon 111.0 152.8 427.3 9,9'40 13,508
+ 26. 84
.
DryweIi if46.6 239.2
+38.7 10,360 13,916
+25.6
!
insulation
+5f.0
Dryweli ISI 51.8 181.6
+71.5 2.8TO 6,2t40 Totals 503.1 811.2
+36.1 4.915 69,507
+25.7 l
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.
.
.
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, i..
.
TABLE 14 Confirmatory Measurement Comparisons for Hatch Nuclear Plant Reported on March 3, 1,9_89,
NRC Licensee Ratio Isotope (uC1/ml)
(uCi/ml)
Resolution (Licensee /NRC)
Comparison H-3 1.9210.06 E-5 2.0E-05
1.04 Agreement Fe-55 2.08 0.06 E-5 2.0E-05
0.96 Agreement Sr-89 8.22 0.35 E-5 8.0E-05
0.97 Agreement Sr-90 4.45i0.18 E-6 4.)E-06
1.06 Agreement Confirmatory M_easurement Com)arisons for Hatch Nuclear Plant Reported on iovember 2E, 1990 NRC Licensee Ratio Isotope (uC1/ml)
(uCi/ml)
Resolu tio_n (Licensee /NRC)
Comparison H-3 6.01d0.24 E-5 5.7E-05
0.95 Agreement Fe-55 4.87d0.19 E-5 3.7E-05
0.76 Agreement Sr-85 7.10t0.28 E-5 4.5E-05
0.63 Disagreement Sr-90 3.2110.13 E-6 2.3E-06
0.72 Disagreement Confirmatory Measurement Comparisons for Hatch Nuclear Plant Reported by Telephone January 2, 1991 NRC Licensee Ratio Isotope (uC1/ml)
(uCi/ml, Resolution (Licensee /NRC)
Comparison Sr-89 7.1010.28 E-5 5.4E-05
0.76 Agreement Sr-90 3.21 0.13 E-6 2.3E-06
0.72 Disagreement Confirmatory Measurement Comparisons for Georgia Environmental Lab Reported on November 12, 1990 NRC Licensee Ratio Isotope (uC1/ml)
(uCi/ml)
Resolution (Licensee /NRC)
Comparison Fe-55 3.70 0.15 E-5 2.6E-05
0.71 Disagreement
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TABLE y CRITERIA FOR COMPARISONS OF ANALYTICAL MEASUREMENTS This attachment provides criteria for the comparison of results of analytical radioactivity measurements.
These criteria are based on empirical
'
relationships which combine prior experience in comparing radioactivity analyses, the measurement of the statistically random process of radioactive emission, and the accuracy needs of this program.
In these criteria, the " Comparison Ratio Limits"1 denoting agreement or disagreement between licensee and NRC results are variable.
This variability is a function of the ratio of the NRC's analytical value relative to its associated statistical and analytical uncertainty, referred to in this program as " Resolution"2, For comparison purposes, a ratio between the licensee's analytical value and the NRC's analytical value is computed for each radionuclide present in a given sample.
The computed ratios are then evaluated for agreement or disagreement based on " Resolution."
The corresponding values for " Resolution" and the
" Comparison Ratio Limits" are listed in the Table below.
Ratio values which are either above or below the " Comparison Ratio Limits" are considered to be in disagreement, while ratio.alues within or encompassed by the " Comparison Ratio Limits" are considered to be in agreement.
TABLE NRC Confirmatory Measurements Acceptance Criteria Resolu lon vs. Comparison Ratio Limits Comparison Ratio Limits Resolution for Agreement
<4 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 2 Comparison Ratio = Licensee Value NRC Reference Value 2 Resolution = NRC Reference Value Xssociated Uncertainty
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