IR 05000312/1987022
| ML20238A482 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 08/11/1987 |
| From: | Cillis M, Yuhas G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20238A462 | List: |
| References | |
| 50-312-87-22, IEC-81-07, IEIN-85-092, IEIN-86-086, IEIN-86-090, IEIN-86-103, IEIN-86-107, IEIN-87-003, IEIN-87-007, NUDOCS 8708310094 | |
| Download: ML20238A482 (26) | |
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U. S. NUCLEAR REGULATORY COMMISSION
REGION V
Report No. 50-312/87-22 Docket No. 50-312 License No. DPR-54 Licensee: Rancho Seco Nuclear Generating Station 14440 Twin Cities Road i
Herald, California 95638-9799
Facility Name:
Rancho Seco Nuclear Generating Station Inspection at:
Clay Station and Sacramento, California
Inspector:
E 8/ /!M7
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M. Cillis, Senior Radiation Specialist Date Signed i
Approved by:
b.9 N h t//a/ W Yllhas, Chief Date~ Signed
G. P.(it4bs Radiological Protection Section Facil
Summary
Inspection on June 15-25, 1987 (Report No. 50-312/87-22).
Areas Inspected:
Routine unanounced inspection by a regionally based inspector of external occupational exposure control and personnel dosimetry, audits and review, liquids and liquid wastes, radiological environmental monitoring, post-accident sampling system, maintaining occupational exposures ALARA, radiation protection and environmental organization and management controls, facilities, followup of licensee reported events, followup items
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identified from previous inspections and a tour of the facility.
Inspection procedures 25544, 25565, 30703, 80521, E4523, 84524, 83722, 83724, 83725,
83727, 83728, 84723, 90713, 92701 and 92/02 were addressed.
Results:
Of the eleven areas inspected, a violation was identified in one Technical Specifications, Section 6.11 (maintaining and adhering to area:
procedures, see section 6).
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DETAILS 1)
Persons Contacted a)
Licensee J. Firlit, AGM, Nuclear Power Production
- B. Croley, Director Technical Services
- D. Keuter, Director, Nuclear Operations and Maintenance J. McColligen, Director Plant Support
- F. Kellie, Radiation Protection Manager
- E. Yochhiem, Chemistry Manager
- W. Wilson, Chemistry Supervisor
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J. Reese, Radiation Health Supervisor I
G. Cranston, Manager, Nuclear Engineering
- J. Robertson, Nuclear Licensing
- R. Colombo, Regulatory Compliance Superintendent D. Martin, Environmental Protection Department Manager
- J. Vinquist, Executive Assistant
- R. Fraser, System Review and Test (PASS)
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- P. E. Turner, Nuclear Training Manager
- R. C. Cherba, Quality Assurance Supervisor E. Bradley, Supervising Health Physicist B. Rodgers, Principal Engineering Technician M. Anderson, M0 VATS Project Manager J. Sullivan, QA Supervisor J. Meyers, QA Supervisor b)
Contractor Personnel 1).
Impell K. Squibbs, PASS System Engineer 2)
United Energy Services (VES)
- G. Marquardt, Superintendent, Health Physics Support and Environmental Monitoring K. Steele, Health Physicist R. Harrington, Health Physicist c).
Nuclear Regulatory Commission i
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- A. D'Angelo, Senior Resident Inspector l
D. Beckman, Consultant Inspector 2)
Quality Assurance Audits and Surveillance An examination was conducted to determine if the licensee's Q^ orogram was consistent with T.S., Section 6.5, " Review and Audit" and with QA and other plant procedures, i
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l A review of audit reports, surveillance reports, procedures and discussions with the licensee's staff were included as part of the
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examination.
It should be noted that the licensee specifically strengthened their Quality Assurance staff in the area of Chemistry, Health Physics and i
Operations as a result of NRC inspection findings identified in i
Inspection Reports 50-312/84-17 and 50-312/84-25.
Personnel having expertise in these areas were hired and an effective audit and surveillance program was initiated starting with the last quarter of 1984 through 1985.
The review of QA surveillance reports related to the licensee's chemistry j
and radiation protection program disclosed there was a significant
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de-emphasis in the performance of QA surveillance.
The de-emphasis was (
noted to start shortly after the beginning of the 1986 season.
At least
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ten surveillance in the area of radiation protection were performed
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between January - April 1986.
None were performed after April 1986.
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number of surveillance in chemistry decreased by 50% in 1986 when compared to 1985 and by another 50% in 1987 when compared to 1986.
Discussions with the licensee's staff confirmed the inspectors observations.
The inspector was informed that the QA surveillance program was decreased shortly after the event of December 26, 1985.
The staff stated that a management decision was made to decrease the surveillance program because resources were required to support activities important for plant restart.
Only a minimum audit and
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surveillance program was maintained to assure compliance with the T.S.
and previous licensee commitments.
A review of Quality Insurance Implement Procedure (QAIP)-6, " Surveillance Oriented QA Surveillance Program" disclosed that scheduling of the QA surveillance program is implemented whenever personnel are on site or otherwise busy in affairs of the Quality Department.
The procedure also includes provisions that allow the Quality Manager to direct personnel to monitor specific areas on a case-by-case real time basis. When fully implemented, the surveillance program prescribed in QAIP-6 provides a good overview of chemistry and radiation protection activities.
The inspector noted that a replacement Quality Assurance Engineer hired f
in August 1986 because of his expertise in radiation protection has not j
been utilized to perform any surveillance or audits in the area of radiation protection to date.
The above observations appear to indicate that the licensee has weakened the QA audit and surveillance program in the area of radiation protection and chemistry.
- The inspector brought the above observations to the licensee's attention at the exit interview.
The importance for maint.aining an effective QA audit and surveillance program was emphasized.
No violations or deviations were identified.
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Internal Exposure Control and Assessment An examination was conducted to determine the adequacy of the licensee's i
internal occupational exposure program to ensure compliance with 10 CFR Part 20.103.
The examination included a review of applicable procedures, personnel exposure records, survey reports, calibration records associated with the calibration of whole body friskers and counters, and personal observations made by the inspector.
The examination also included the review of processing and control of respiratory protection equipment.
The use of respiratory equipment was found to be consistent with 10 CFR 20.103 and NUREG 0041, " Manual of Respiratory Protection... Materials."
The inspector noted that the licensee does not have any permanent facilities for processing respiratory equipment.
The licensee currently utilizes a contracted unit specially designed for processing respirators (see section 4).
I Procedures for assessing individual intakes of radioactive material were l
being implemented and no abnormal results related to. internal dose assessment were identified from the review of personnel exposure records.
Particular attention was given to reported personnel contamination i
occurrences.
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A review of whole body counting maintenance and calibration records disclosed there was an increase in the number of maintenance items requiring closer attention.
The plant Health Physicist was aware of the i
problems and had submitted a request to his management to replace the current whole body counter with a new one and to obtain a " quickie" whole body counter (see section 4).
This observation was diccussed with the licensee's staff attending the exit interview.
A new respirator fit-up booth (see section 4) had just arrived at the l
time of this inspection.
The plant Health Physicist stated that the old fit up booth will be used to test respirator filters.
Personnel skin and clothing contamination occurrences for 19PS and 1987 were reviewed.
The occurrences were well documented and all contamination occurrences were evaluated for probable cause and for trending purposes.
Additionally, where appropriate, licensee dose assessments resulting from skin contamination occurrences were performed.
The inspector noted that licensee procedures do not require dose assessment evaluations resulting from clothing contamination occurrences.
This was brought to the licensee's attention during the inspection.
The Radiation Protection Manager informed the inspector that plant procedures will be revised to include a requirement for performing dose assessments resulting from clothing contamination occurrences.
The review of personnel skin and clothing contamination occurrences did not reveal any significant levels personnel exposures.
The inspector was informed that two new whole body friskers were recently received on site (see section 4).
The friskers will be installed at the control point exit from the auxiliary and reactor building.
The friskers
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are expected to improve the licensee's external and internal exposure control program.
No violations or deviations were identified.
4)
Facilities and Equipment An examination was conducted to determine.if any new changes have been made or were planned in the area of radiation protection.
The examination disclosed the following changes that were in place, in progress or under consideration:
A new Regenerant Holdup Tank (RHUT) of 100,000 gallon capacity was under construction.
The new RHUT, designated as "C" RHUT (T-950C),
was deemed necessary for the purpose of separating potentially contaminated from non-contaminated secondary waste water, increasing the plants processing capabilities and flexibility, and to provide additional cleanup components that will be used to maintain off-site liquid effluent discharges within 10 CFR Part 50, Appendix I limits.
The "C" RHUT will be used for processing non-radioactive liquid effluents.
A new sluiceable demineralized system to clean up 'A'
and 'B'
RHUT water prior to transfer to the retention basins is under i
construction.
J A make-up demineralized waste sump, two sump pumps, and oischarge header from the pumps were under construction.
The new make-up demineralized will be lined with high density polyethylene.
Make-up demineralized regenerate wastes and other non-radioactive wastes will be routed to the new sump.
A new Dynatech respirator fit-up booth had just arrived on site at the time of this inspection.
Arrangements were underway to contract the services of a " quickie" whole body counting system and a new laydown whole body counting system.
Two new Eberline, Model PCM-1B whole body friskers had just arrived on site at the time of this inspection.
Arrangements were underway to review the contract for leasing services of a self contained trailer that is specifically designed for processing respiratory equipment.
No violations or deviations were identified.
5)
External Occupational Exposure Control and Personnal Dosimetry The licensee's personnel dosimetry program and control of external occupational exposure was examined for the purpose of assuring compliance-with:
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10 CFR Part 19.13, " Notifications and reports to individuals" 10 CFR Part 20.101, " Radiation dose standards for individuals in restricted areas" l
10 CFR Part 20.102, " Determination of prior dose" 10 CFR Part 20.104, " Exposure of minors" l
10 CFR Part 20.202, " Personnel monitoring"
.10 CFR Part 20.401(a),'" Records of surveys, radiation monitoring and'
disposal" i
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10 CFR Part 20.408, " Reports of personnel monitoring on termination
of employment or work" Licensee procedures that have been established for assuring compliance with the above requirements and selected personnel exposure records were reviewed and discussions were held with the plant Health Physicist and
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his staff.
The examination disclosed that the licensee utilizes a personnel dosimetry service offered by a vendor certified under the National Voluntary Laboratory Accreditation Program (NVLAP).
The licensee uses a backup Panasonic Thermoluminescent Dosimeter.(TLD) Service to provide a quick assessment of personnei ' exposure while performing critical work in high radiation areas.
The licensee's program also includes provisions-for extremity monitoring, multi-whole body monitoring and for neutron l
dose assessments.
Additionally, quality control activities are included in the program.
The inspector verified that the panasonic TLD' reader was routinely
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calibrated in accordance with established procedures.
The inspector also reviewed a licensee training program that is provided to the dosimetry staff.
o The Plant Health Physicist (HP) informed the inspector that the confirmatory measurements program identified anomalies, in the range of 130%, with the vendor supplied TLD programs' capability for measuring non penetrating radiation.
The supervisor stated that for the most part the results were on the low side of the spectrum, e.g. minus ten to I
twenty per cent.
The licensee ascertained that this anomaly is attributed to the differences used in calibrating the TLD's.
For instance, the vendor uses a Sr-90 source as a standard for interpreting shallow dose equivalent; whereas, the licensee' utilizes a Depleted Uranium (DU) slab.
The licensee's evaluation of this matter has included ongoing discussions with the vendor.
This item will be examined during a subsequent inspection (87-22-03).
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system; however, the anomalies were much greater, e.g. up to 260%.. The plant HP stated that the vendor supplied TLD service is used as the
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official record for assigning personnel exposures.
He added that the Panasonic system is being used as a backup to verify the results obtained from pocket ionization chambers.
The HP stated.that the Panasonic TLD system has not been properly calibrated to measure the response to non penetrating radiation.
He added that a long term action is underway to first establish the average beta. energy from the radionuclides in the waste streams.
The licensee stated that they will attempt to resolve the problem associated with the Panasonic systems capability to accurately measure the response to non penetrating radiation.
This item will be examined during a subsequent inspection (87-22-04).
The inspection revealed that a program to compare (i.e. trend) pocket ionization chamber and TLD results had just been implemented.
Additionally,' procedures were being developed for the purpose of establishing an acceptance criteria for the personnel dosimetry l
confirmatory measurements program.
The Radiation Protection Manager informed the inspector that provisions have been made with the dosimetry vendor to perform a survey of the neutron energy spectrum inside the containment while the reactor is at full power.
The study will be performed after plant restart.
No recent audits or surveillance of the licensee's dosimetry program were performed.
The Radiation Protection Manager was aware of this and
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had taken action by requesting an internal audit of the program to be performed by the licensee's Environment and Health Physics Support group.
The audit was scheduled to commence during the week starting June 29, 1987.
No violations or deviations were identified.
6)
Maintaining Occupational Exposures ALARA A).
General Information An examination was conducted to determine if the licensee's ALARA program was being effectively' implemented.
It should be noted that previous NRC inspections have found weaknesses and inconsistencies with the licensee's implementation of their ALARA program.
These inconsistencies are documented in Inspection Reports 50-312/84-17, 50-312/85-03, 50-312/65-06, 50-312/86-11 and 50-312/86-16.
Similar weaknesses were also identified in SMUDs audit report 86-004.
B). Management Policy and Commitments The Introduction in the licensee's ALARA Manual, dated December 14, 1984 states:
" Sacramento Municipal Utility District management is committed to a policy af maintaining radiation exposures As Low'As Reasonable Achievable (ALARA) (See Appendix A-1)."
The Appendix A-1 ALARA policy statement endorsing the introductory statement implementing an effective ALARA program was signed by the
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top four levels of SMUD's management organization.
The operation's ALARA philosophy described in the ALARA policy statement states that
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the guidelines in Regulatory Guides (R.G.) 8.8 and 8.10 shall be
applied to the extent practicable through the planning, design,
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construction, operating, maintenance and decommissioning of the
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plant.
l Rancho Seco Management Process Manual (RSMP)-309, Rev. O, dated April 17,1987, entitled "ALARA Program" includes such statements as:
"A training program will ensure that all Rancho Seco personnel are trained in their responsibility toward the ALARA Program."
"The ALARA program applies to all activities which may lead to exposure for workers or the general public.
The ALARA concept requires active involvement and commitment from all levels of management, supervision, and labor."
This ALARA program must meet NRC requirements and the intent of NRC guides as defined in the cited references,...."
Staffing and training must be adequate to communicate and implement the ALARA program."
l The ALARA Manual shall identify personntl training to comply i
with NRC requirements and guides."
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"The Manager, Quality Assurance is responsible for performing
audits of the ALARA Program and ALARA data base."
j Additionally, Article 1 of Radiation Control Manual (RCM) procedure AP 305, " Foundation for the Rancho Seco Radiological Controls j
Program" states:
" Specific details as to how the Rancho Seco RPP is i
implemented are promulgated in the Radiation Control Manual parts II
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and III and the Rancho Seco ALARA program as defined in the ALARA Manual." Article 1 also states:
"The Radiological Protection Program in conjunction with the Rancho Seco ALARA program will
assure that personnel exposures are maintained ALARA." The RPM goes j
on to state:
"There is no person or group involved in operations of i
Rancho Seco who does not have some degree of responsibility for the Radiation Protection Program and ALARA Programs.
Failure of any person to recognize this responsibility or to comply with issued procedures will not be tolerated."
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The above quotes provide a very small part of SMUD's commitment for maintaining an ALARA program.
It should be noted that Technical Specifications, Section 6.11, " Radiation Protection Program" states:
" Procedures for personnel radiation protection shall be prepared j
consistent with the requirements of 10 CFR Part 20 and shall be j
approved, maintained and adhered to for all operations involving personnel radiation exposure."
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C).
ALARA Program Implementation Section 3.3.3 of the ALARA Manual assigns the Supervising Health Physicist everall responsibility for the administrative duties of the District's ALARA program.
Section 3.3.3 assigns the Supervising Health Physicist with responsibilities covering fourteen different areas.
The Supervising Health Physicist informed the NRC inspector at the onset of the inspection that he has not performed the functions of the ALARA coordinator for the past four to six months.
He added that his involvement prior to that time was on a part time basis.
The supervising health physicist stated his participation in the ALARA program was influenced by his involvement and responsibilities in other activities such as the Environmental Monitoring program.
The Supervisor asked to be dismissed from the NRC inspection associated with the ALARA program.
Section 2.9 of the ALARA Manual states:
The Quality Assurance Department audits the performance of the radiation protection program and the ALARA program requirements and reports the audit results directly to the Management Safety Review Committee.
These audits shall include records of procedure reviews, past exposure records, plant inspections, consultation with the radiation protection staff or outside consultants, and steps that have been taken to reduce exposures.
The examination disclosed that Quality Assuranco audits of the ALARA program were performed.
However, the audits did not include a review:
(1) of the ALARA Manual implementing procedures, (2) past exposure records, (3) plant inspections or (4) steps taken to reduce exposures.
The manual also states the audits should be conducted formally and periodically.
Discussions with the Quality Assurance staff disclosed they were unaware of the scope of the. audit requirements specified in the ALARA Manual.
Section 2.2 of the ALARA Manual establishes the following ALARA training program:
" Training (see Section 4.0) is provided to familiarize all station personnel with the basic principles of sound ALARA practices and with District management's commitment to ALARA.
This training is tailored to the duties and responsibilities of those receiving the instruction, and is structured to give guidance on the mechanisms of maintaining the individual's exposure ALARA.
Refresher training is provided annually.
"A detailed presentation of ALARA engineering practices is provided to engineering personnel to reinforce their knowledge and keep it current.
The information presented is reviewed periodically and modified where necessary, to reflect current techniques and adjustments based on experience in operations and engineering that affect radiation exposure._
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"ALARA Training is required for all personnel whose duties routinely require:
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"1) unescorted access to the Restricted Area; i
"2) vorking with radioactive materials or equipment;
"3) entering radiation areas;
"4) directing the activities of others who enter radiation areas or who work with radioactive materials or equipment;
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"5) the engineering and design of facilities, systems, and j
components at Rancho Seco Nuclear Generating Station."
Sections 4.0, 4.3 and 4.4 of the ALARA Manual states:
"4.0 TRAINING
"A training program ensures that all on-site personnel understands management's commitment to ALARA, workers'
rights, ALARA concepts and radiation protection rules and regulations.
The scope and depth of training is appropriate for the individual's job.
The effectiveness of training is evaluated during the course of instruction to assure knowledge and understanding of subject material.
Refresher training is required annually to maintain an appropriate level of knowledge and performance.
"4.3 Supervicor/ Foreman Training
"ALARA subjects are incorporated in annual training presented to SMUD and contractor supervisors and foremen who are or may be involved in controlled area work.
SMUD and contractor health physics personnel also attend this training.
ALARA subjects include:
" Task preplanning techniques including review of previous experience, identification of crew and equipment requirements, scheduling, crew training and work area preparation.
" Task performance including supervision techniques, radiation work permits, dosimetry use, and data collection.
" Post task reviews including crew debriefing, evaluation of job experience, and documentation of data.
" Review of applicable ALARA and radiation protection procedures.
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"4.4 ALARA Engineering Training
"This training is presented to engineers and draftsmen who design Station systems, or changes to those systems or plant components.
Subjects include:
" Applicable regulations and regulatory guidance.
" Engineering and design exposure reduction considerations.
" Engineering and design ALARA procedures."
The examination disclosed at least ten supervisors and foremen who had not received the ALARA training.
The only group that appeared to support the ALARA training program was the radiation protection staff.
Section 2.2 and Section 4.4 require that a detailed ALARA annual training program be provided to engineering personnel and to draftsman who design station systems, or changes to those systems or plant components.
The purpose for the training is to reinforce their knowledge and keep it current.
The training provided to approximately 25 to 30 indivi. duals from the licensee's engineering staff who had signed Engineering Change Notices and Design Basis Reports for changes made to plant components were examined.
The examination disclosed that only two individuals had received the required training, six had not received any ALARA training since 1985, and the remaining seventeen individuals had not received the training at all.
Section 2.3 1 of the ALARA manual states that the cognizant engineer is responsible for establishing a man-rem estimate for the ECN (Engineering Change Notice) at the time a Design Basis Report (DBR)
is required, and for ensuring that ALARA design features have been adequately considered, etc.
Section 2.3.1 goes on to state that the j
Nuclear Engineering Health Physics staff will assist the cognizant
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engineer in establishing the man-rem estimates, etc.
The inspection disclosed that all man-rem estimates are performed by i
the Principal Engineering Technician (PET) and his staff.
The PET is a member of the ALARA group who normally acts as the ALARA
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coordinator whenever the permanently assigned ALARA coordinator is
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not available.
Nuclear Engineering Procedure (NEP)-4109, " Rancho Seco Configuration Control Procedure" Section 2.3 states that the Nuclear Engineering Department is responsible for ensuring ALARA review of design i
modifications is performed per the ALARA progro." The ALARA manual
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also requires that ECN's receive an ALARA review.
The examination disclosed that there is no way to verify that all required ECNs receive an ALARA review.
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ALARA Considerations for the MOVAT's Project Discussions held with the PET and the current Project Manager associated with the refurbishment of Motor Operated Valves Analysis
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Tests (MOVATs) revealed that the job planning and scoping was not l
consistent with the ALARA concept.
This finding is also substantiated in section 4 of Region V Inspection Report 50-312/87-09.
Original man-rem estimates of less than 18 man-rem were established when the project was first started in the second half of 1986.
The current man-rem estimates had exceeded 63 man-rem at the conclusion of this inspection, with approximately 75 to 80% of the work completed.
The PET and current MOVAT Project Manager informed the inspector j
that the initial job scoping and planning was extremely poor.
The j
inspector reviewed at least nine memos from the radiation protection staff that also substantiated the need to perform additional and/or rework activities on the MOVAT's project.
SMUD memorandum dated February 23, 1987 attributed the need for redetermining M0 VATS man-rem estimates because of an anticipated q
re-work of at least fifty percent of the M0V's that were originally considered to be complete.
The letter states that the Health Physics and Environmental Programs Support Group had repeatedly experienced difficulties in obtaining information and in being advised of problems encountered with the M0 VATS task.
The report also identified that there was a lack of communication between the MOVATs and ALARA groups.
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Other Observations
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l The examination disclosed that no Quality Assurance audits or surveillance were conducted to determine the effectiveness of the licensee's ALARA program during the performance of critical work performed in areas such as Steam Generator Sleeving Project, M0 VATS, Pressurizer Heater Bundles, Under Vessel In-Service-Inspection, Snubbers, Steam Generator Eddy Current, In Service Inspection (ISI), and Reactor Coolant Pump Repairs.
Section 2.3 of the ALARA Manual requires a Management Safety Review Committee (MSRC) review of all tasks with a man-rem estimate of 50 man-rem or greater.
The examination disclosed that an MSRC review had not been conducted for the M0 VATS project even though 63 man-rem had been expended to date.
Remaining ALARA functions prescribed in the ALARA Manual appeared to be consistent with the Radiation Control Manual and ALARA Manual.
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i A 416 Person-Rem goal has been established for 1987.
For 1986, the initial estimate was set at 165 Person-Rem; however, by the
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end of 1986 a total of approximately 505 Person-Rem had been
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expended.
Discussions with the PET revealed that the 1986 work i
scope changed several times after the original estimate was l
made.
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ALARA Manual Review
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l ALARA Manual procedure ALARA-1, "ALARA Manual Control" assigns the
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Supervising Health Physicist the responsibility for reviewing the j
ALARA Manual every two yeat s.
A revision to the ALARA Manual was made after a review was conducted
at the end of 1986.
The revised manual was submitted to the SMUD
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staff and MSRC for review and approval shortly after the revisions were made.
The revised manual had'not been issued as of July 1, 1987.
Initially, the revised manual assigned specific responsibilities to ensure the ALARA concept was consistent with R.G. 8.8 and 8.10.
After several recent reviews, the manual includes the considerable use of " permissive" terms.
The manual now lacks specificity and will be open for interpretation if it is approved for issue.
It appears that this will lead to a further degradation of the ALARA program.
G).
Finding The above observations were brought to the attention of licensee at the exit interview.
The inspector informed the licensee that failure to implement the ALARA Manual (see section 6(B), above) was an apparent violation (87-22-01).
The licensee acknowledged the inspectors observations and agreed to notify the inspector of their plans to resolve the observations.
The inspector was contacted by the Manager, Technical Services and a
representative from Licensing on July 2, 1987 to discuss plans to
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resolve the issues.
The inspector was informed of the following:
The PET will be temporarily assigned as the acting ALARA
Coordinator.
j The current ALARA Manual will be enforced until the time that the revised manual is issued.
A goal of August 15, 1987 has j
been established to issue the revised manual.
The use of
" permissive" terms will be eliminated prior to issue so that it is auditable.
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The ALARA training program will be redefined and is expected to
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be implemented by the end of 1987 or early part of 1988.
NEP-4109 will be revised by the end of July 1987 to clarify the processing of ECNs/DBRs for ALARA purposes.
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Remaining observations will be addressed accordingly.
7)
Control of Radioactive Materials and Contamination, Surveys, and Monitoring An examination was conducted to determine if radioactive materials and contamination were effectively controlled and if adequate surveys were performed.
Applicable procedures and survey records were reviewed and direct observation of contamination control practices were made during tours of the licensee's facilities.
The licensee's Radiation Protection is responsible for monitoring all materials leaving controlled areas and for controlling all radioactive materials.
The monitoring practices observed appeared to be consistent with IE Circular 81-07 and IE Notice 85-92.
Licensee. procedure AP 305-9A, " Removal of Tools and Equipment from Controlled Areas", provides the instructions for assuring all materials released from controlled areas are monitored in accordance with the guidelines recommended in the Circular and Notice.
A review of routine surveys performed within the protected area, but outside the controlled areas, disclosed that all facilities within the protected area are monitored at least monthly.
The review of survey records associated with activities performed within the Reactor Building, Auxiliary Building, Radwaste Systems, Turbine Building and those requirect by the licensee's Radiation Work Permits disclosed that the licensee's monitoring program was consistent with 10 CFR Part 20.201, " Surveys" and 10 CFR Part 20.401, " Records of Surveys, Radiation Monitoring, and Disposal."
No violations or deviations were identified.
8)
Annual Radiological Environmental Operating Report An in office review of the 1986 annual radiological environmental operating report was conducted.
The following observation was made:
It should be noted that the tabulated detection limits for air particulate, as shown in Appendix B, Table 1, has several anomalous limit values that should be corrected (e.g. Ru-106, Cs-137, Co-58, Mn-54).
This was brought to the licensee's attention during the inspection.
The licensee's staff stated they would look into the matter.
This item will be examined during a subsequent inspection (87-22-02).
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Tour (
A tour of the auxiliary building, outside storage areas, basin area and j
the new low-level waste storage facility was performed during the inspection.
The inspector made independent radiation measurements using an Eberline R0-2 portable ion chamber survey instrument, S/N 2694, that was due for calibration on July 23, 1987.
Licensee posting and labeling practices were confirmed to be consistent with the regulatory requirements prescribed in 10 CFR Part 19.11,
" Posting of Notices to Workers" and 10 CFR Part 20.203, " Caution Signs, Labels, Signals and Controls."
Cleanliness in the Auxiliary and Reactor Buildings was in need of attention.
An excessive amount of splintered plywood, two by fours, and l
wood scaffolding were observed during the tour.
Contamination control
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boundaries on the -20 foot level of the Auxiliary had expanded rather considerably from the observations made during the previous inspection.
The above observations were brought to the licensee's attention during j
the tours and at the exit interview.
The licensee's staff took immediate J
action in the removal of splintered wood.
The licensee's staff infomed the inspector that arrangements were being made to condense the
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contamination control boundries to locations that were established prior to the outage.
The staff indicated that the boundaries had been j
established to support M0 VATS and other related work.
I Additional observations are as follows:
Work practices were consistent with instructions provided on Radiation Work Permits.
j Portable instruments used for air sampling and radiation detection were in current calibration.
No unmonitored personnel were observed in the areas toured.
No violations or deviations were identified.
10)
Liquid and Liquid Wastes A)
An examination of the licensee's program for control and management of liquids and liquid wastes was conducted for the purpose of verifying compliance with the regulatory requirements prescribed in the following Sections of the Technical Specifications (TS).
T.S. Section Title 3.1.4 Reactor Coolant Activity 3.10 Secondary System Activity 3.15 Radioactive Liquid Instrumentation L
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3.17.1 Liquid Effluents 3.17.2 Dose 3.17.3 Liquid Holdup Tanks 4.0 Surveillance Standards 4.19 Radioa.tive Liquid Effluent Instrumentation 4.21.1 Liquid Effluents 4.21.2 Doses
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4.21.3 Liquid Holdup Tanks l
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6.5.2.8.(k),(1)
Audits
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6.8 Procedures 6.9.2 Reports i
6.16 Offsite Dose Calculations Manual Matters related to this subject are discussed in sections 2, 3 and 5 of this report.
The examination also include the review of applicable procedures, proposed license amendment changes, personnel observations by the inspector and discussions with the licensee's staff.
The findings documented in Inspection Reports 50-312/86-15 and 50-312/86-23 were
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also reviewed and used as a guideline-for performing the
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examination.
B).
Audits The inspector verified that audits of the environmental monitoring program and the Offsite Dose Calculation Manual (ODCM) were performed in accordance with the requirements prescribed in T.S.,
Sections 6.5.2.8(k) and (1).
Rancho Seco Audit Reports 0-811 dated June 18, 1986 and 0-814 of July 1986 were reviewed.
The review disclosed that the audits did not appear to include an in-depth review of plant procedures.
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l Discussions with the licensee's staff disclosed that an audit of the I
licensee's liquid and liquid waste program was scheduled for the week of June 29, 1987.
The audit was to be performed by an independent offsite contractor.
The inspector noted that the audit started on June 25, 1987.
No violations or deviations were identified.
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Changes
I Changes observed in the liquid waste processing system is discussed in sections 4 and 11 of this report.
This area.will be examined during a subsequent inspection.
The Offsite Dose Calculation Manual was in the process of being revised for the purpose of including the comments identified in Inspection Report 50-312/86-15 and to address the Franklin Research Center Report findings that were reported to the NRC, and to
incorporate the comments from the May 5, 1987 meeting in the Region j
V Office.
The revision is expected to be completed by August 1,
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1987.
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Additional changes include a complete revision of the Technical Specifications.
The proposed changes are to be submitted as Proposed Amendment 155.
The proposed changes were submitted to NRR at the end of June 1987.
Discussions with the licensee's staff disclosed that the Updated Safety Analysis Report (USAR) was in the process of being revised, j
The revision is expected to include changes that have occurred in
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the control of liquid and liquid wastes.
No violations or deviations were identified.
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D).
Liquid Waste Releases A total of 27 liquid waste releases were recorded for the period of
January 1 through June 15, 1987.
Sample counting times for each of the releases were from 4000 to 6000 seconds.
Typical Lower Limit of Detection (LLD) vales for Cs-134, Cs 137 and Co-60 seen were <1.7E-8 pCi/ml, <2.7E-8 pCi/ml
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and <3.02 E-8 pCi/ml, respectively.
j The 27 releases totaled 2.6031E6 gallons prior to dilution.
The following table summarizes the results of gamma scans for the twenty-seven releases in which radioactive material was identified:
j Release No.
Tank Result pCi/ml 87-002 A-RHUT Cs-134 3.73 E-8 Cs-137 9.44 E-8 87-005 8-RHUT Cs-134 5.06 E-9 Cs-137 1.29 E-8 87-008 A-RHUT Cs-137 4.08 E-08 87-010 A-RHUT Cs-137 2.14 E-08
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)87-014 A-RHUT Cs-137 4.61 E-8 87-025 B-RHUT Cs-60 1.28 E-8 Cs-137 4.16 E-8 i"87-026 B-RHUT Cs-60 1.7 E-8 Cs-137 5.54 E-8 87-027 A-RHUT Cs-137 5.85 E-8 l
The maximum hypothetical doses calculated as a result of liquid j
releases completed up to May 11, 1987 were reported as follows:
j Adult whole body
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1.22 E-1 mrem Teenager whole body -
1.19 E-1 mrem Child whole body
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1.88 E-1 mrem l
No violations or deviations were identified.
11) Followup Items An examination was conducted for the purpose of determining the status of corrective actions taken by the licensee in resolving previous inspection findings, Licensee Event Reports (LERs), Part 21 Reports, IE Information Notices (ins), and other non-routine events reported by the licensee.
The examination disclosed the following:
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Information Notices
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(Closed) Information Notices (IN 86-86, IN 86-90, IN 86-103, IN 86-107, IN 87-03 and IN 87-07):
The inspector verified that the licensee's staff received and evaluated the following ins:
IN No.
Title IN 86-86 Clarification of Requirements for Fabrication and Export of Certain Previously Approved Type 'B'
Containers IN 86-90 Requests to Dispose of Very Low-Level Radioactive Waste Pursuant to 10 CFR 20.302
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l IN 86-103 Respirator Coupling Nut Assembly Failures
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i IN 86-107 Entry into PWR Cavity with Retractable Incore
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Detector Thimbles Withdrawn
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IN 87-03 Segregation of Hazardous and Low-Level Radioactive Wastes IN 87-07 Quality Control of Onsite Dewatering /
Solidification Operations by Outside Contractors
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This matter is closed (IN 86-86, 1H 86-90, IN 86-103, IN 86-107, IN i
87-03 sad IN 87-07).
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B)
Enforcement Item j
(Closed) 87-05-03 - Tile inspector verified that the corrective actions identified in the licensee's timely response to this item were completed.
This item was associated with a radioactive material shipment.that was found to have a hot spot and had not been properly labeled.
Corrective actions included:
Revision (s) of Radwaste Manual procedures Training of involved personnel
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The revisions include provisions for assuring that radioactive materials packaged for shipment are carefully scanned with appropriate radiation detection instruments and that all packages are properly labeled in accordance with Department of Transportation requirements.
Thit, matter is closed 87 05-03.
C.
FollowupJtems (0 pen) 8'F16-12 - Inspection Report 50-312/83-16 identified that Plant EffToent Padiation Monitbring (PERM), R15044, did not include orovisionc ! tor' monitoring both the containment atmosphere and reactor building vent as required by NUREG 0737, Item II.F.1.
The monitor only sampled tha containment atmosphere.
This condition was determined to be acceptable by NRR since the reactor building purge valves are locked closed except at cold shutdown or under refueling conditkors.
The dantainment equalizing valves would also remain
shut during reactor power operations.
Modifications and are being made to the reactor building monitor, R15044., and auxiliary building monitor, R15045, to meet the NRC requirements for operation of the containment equalizing valves at power.
The modifications which are currently scheduled for completion prior to plant restart should provide a radiation monitoring system which meets all of the requirements of NUREG 0737,
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i Item II.F.1.
The modifications are being performed under Engineering Change Notice (ECN) 0193.
This item will be re-inspected upoh completion of the modifications.
L (Closed) 26-1b,-02 - Inspection Report 50-312/86-15 identified that the licensee was unable to locate records associated with liquid erfluents-released frow t5e "A" Regenerative Holdup Tank on October
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\\23, 1985 and November 13,N1985.
The release permit numbers were 85-203 and 85-213, respectively.
This was identified as an example of failure to comply witn Technical Specifications (T.S.), Section 6.10.2.
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AnexaminationdisclosedtItatthemissingrecordswerefoundand placed in the license 6s permanent files.
A review of the missing
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records did not disclose any gamma peaks above the LLD values.
This matter is closed (86-15-02).
(Closed) 86-15-06 - Inspection Report 50-312/86-15 raised the following questions:
The acceptability of utilizing connections of a non quality class (e.g. plastic piping) to transfer liquid effluents from the Demineralized Reactor Coolant Storage Tank (DRCST) to the A or B RHUTs.
Maintaining the quantity of radioactive material contained in all outside storage tanks to 10 curies if they are not
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surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste j
treatment system.
l An examination disclosed that the temporary plastic piping is being l
replaced with a stainless steel transfer line.
Additionally, a review of Proposed License Amendment Number 155 disclosed that the licensee is requesting a revision to T.S. Section 3.17.3, Liquid Holdup Tanks.
The licensee has requested that the Borated Water Storage Tank (BWST), Demineralized Reactor Coolant Storage Tank (DRCST) and Miscellaneous Water Holdup Tank (MWST) be included under T.S., Section 3.17.3.
This matter is closed (86-15-06).
(Closed) 86-15-12 - Inspection Report 50-312/86-15 identified that the licensee's Semiannual Radioactive Effluent Report, dated March 3, 1986, contained several errors.
An examination disclosed that all of the errors were corrected.
This matter is closed (86-15-12).
(0 pen) 86-37-03 - Inspection Report 50-312/86-37 raised a question as to ' nether radiation monitors R15701 and R15702 would provide representative sampling of the Control Room - HVAC airstream.
The subject monitor detectors are installed approximately ten to
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eighteen feet from the CR-HVAC air handler intakes.
An examination of this item was conducted during the inspection.
The results of the examination is discussed in Region V Inspection Report 50-312/87-21.
This matter will be examined during a subsequent inspection (86-37-03).
(0 pen) 83-16-02, 84-01-02, 35-28-11 The status of these long-term I
improvement items was examined.
The importance for completing these
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items in a timely manner was brought to the licensee's attention in Inspection Reports 50-312/86-11 and 50-312/86-37.
Actions to improve item 83-16-02 started in 1979, and actions to improve items 84-01-02 and 85-28-11 started before 1982 and July 1983, l
respectively.
The inspector was informed on several occasions that
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the work on these items would be completed by January 1, 1987 and by
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r-O ao later than plant startup which was tentativel.y scheduled for March or May 1987.
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b The examination disclosed that although some progresh ad been made since the previous inspection; the items are still iticomplete.
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Another individual has boen assigned responsibility for these items.
The individual was unable to state when the items would be completed i
other than to state possibly prior to plant startup.
The above observations were brought to the licensee's attention at the exit interview.
The inspector re-emphasized previous concerns raised in Inspection Reports 50-312/86-11 and 50-312/86-37.
These items will
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be examined during a subsequent inspection (83-16-02, 84-01-02 and 85-28-11).
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(0 pen) 84-27-01 - Inspection Report 50-312/84-27 identified that the construction of a new Low-Level Radioactive Waste Storage Facility designed and approved to provide storage of low-level, non-liquid radioactive waste was nearing completion.
An examination was conducted to determine the status of the facility.
The examination disclosed that the preoperational testing phase was nearing completion and that operational procedures were also nearing completion and that training has been provided to the operating staff.
A review of the licensee's Safety Analysis, ECN A-3646, Revision 1, dated March 13, 1987 was conducted.
The following observations were notec;.
a)
The revised Safety Analysis report extended the utilization of the facility from what was approved during the original Safety Review which was performed prior to the construction of the facility.
The original Safety Analysis provided authorization for the storage of low-level radioactive wastes.
Revision 1 of Safety Analysis extend the use of facility to include the following types of activities:
Sorting and segregation of Dry Active Waste (DAW).
- Consolidation of uncompactable waste prior to shipment Laundry inspection, sorting, survey and storage.
i Respiratory cleaning unit operation DAW compaction (box comprStor)
Decontaminatir9 of equipment and/or volume reduction of fixed contaminated equipment and gear.
Encapsulation of solid radioactive materials.
(b)
It should be noted that the heating, ventilation and air conditioning air exhaust from the facility is HEPA filtered and
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discharges to the atmosphere via a common release point.
The airborne effluents will be monitored continuously for.
particulate, gases, and iodine via an isokinetic probe.
Portable engineered ventilation units equipped with HEPA-filters will be used to supplement the; normal' ventilation system.
The normal ventilation system'is designed to close in the event of an alarm from 'le monitor.
The alarm sounds locally and in the control room.
(c) The licensee's revised Safety Analysis report determined that:
The use of the facility would not increase the probability of occurrence or_the consequence of an accident or malfunction of equipment important to safety that were previously evaluated in the original Safety Analysis report.
The use of the facility would not create the possibility for an accident or malfunction nf a different type than previously evaluated in the original Safety Analysis report, namely radiological exposure.
The activities planned in the facility will not result.in a release of airborne radioactivity to the environment and shall not require a Technical Specification amendment.
(d) The Safety Analysis report did not state whether necessary.
changes would be made to the applicable sections of the USAR.
Based on further review of the licensee's evaluation, 10 CFR Part 50.59, T.S. requirements and USAR commitments; the following anomalies were noted:
10 CFR Part 50.59(a)(1) states in part, that the licensee may make changes to the facility anc' operations as described in the Safety _
Analysis Report without piv approval, provided a change in the Technical Specifications is not involved.
The Bases to the Technical Specification, Section 4.12,
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" Auxiliary Building and Spent Fuel Building Filter System,"
lists the Radwaste Area as being a potential release pathway.
j It should be noted that releases from the dry' active waste
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compactor located.in the Radwaste Area are made via the
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Auxiliary Building and Spent Fuel Building Filter System.
It would appear that the same requirement would also apply to the new facility.
- Neither the Technical Specifications or the Updated Safety-~
Evaluation Report show:
(1) the new low-level radwaste storage facility, and (2) the HVAC and associated radiation monitoring system.
j The need to revise the Technical Specifications to include the new low-level radwaste facility as a potential release pathway was
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brought to the licensee's attention at the exit interview and was subsequently discussed with the NRR Project Manager during the week of June 29, 1987.
Resolution of this matter will be documented in a subsequent report (84-27-01).
(Closed) 87-05-02 - Inspection Reports 50-312/86-05, 50-312/86-06 and 50-312/87-05 identified some concerns with the licent.ee's radiation protection and the environmental monitoring and health physics organizations.
Additional concerns for resolving health physics issues and improving communications were addressed in the listed inspection reports and other inspection reports that were issued since the event of December 26, 1985.
Inspections conducted since the December 26, 1985 event reflect a general improvement in the licensee's chemistry, radiation protection and environmental monitorir.g and health physics support organizations.
These improvements have lead to improvements in communications and in the resolution of health physics issues.
Discussions with the Radiation Protection Manager and the newly appointed Environmental Protection Manager (EPM) disclosed that the concerns raised in Inspection Report 50-312/87-05 were adequately resolved.
The EPM informed the inspector that he had the authorization to fill nine slots in the Environmental and Health Physics Support group.
The EPM stated that his goal was to fill the vacancies as quickly as possible.
Discussions with the RPM revealed similar managemer,t support in assuring that his organization is adequately staffed to support plant activities.
The inspector noted that improvements in the area of communications are continuing on an upward trend since the health physics group assumed full control for radiation protection functions.
The RPM and Chemistry Manager indicated that bccause of the split of the chemistry and radiation protection group in 1985 and the increased staffing; they are now able to devote their resources to resolving issues in their respective areas of responsibility.
The inspector discussed the above observations at the exit interview.
The importance for fully implementing the Environmental Monitoring and Health Physics Support organization that was j
presented to the Region V staff during the Enforcement Conference of
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June 1986 was re-emphasized.
This matter is closed (87-05-02).
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(0 pen) 86-37-01 - Inspection Reports 50-312/86-37 and 50-312/87-05 provide a summary on the status of the licensee's Post A;cident Sampling System (PASS).
The reports identifiea that modifications were being made to the PASS for the purpose of improving reliability.
The report also identified that the modifications have been in progress since March of 1986 and that system testing will be observed by the NRC's Region V staf _-__ - _.
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An examination was conducted for the purpose of determining the status of modifications and system testing.
The following observations were made:
j Surveillance Test Procedure (STP)-432 for containment atmosphere sampling was initiated during the first week of May 1987.
Problems with the sample pump were encountered.
The pump problems were resolved on or about June 20, 1987, and tests for obtaining a containment atmosphere sample were resumed.
A problem with the containment atmosphere sample line was identified shortly after resuming the tests.
Discussions with i
the licensee's staff disclosed that the containment atmosphere
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sample line may be clogged.
The problem with the sample line was still unresolved as of July 2, 1987.
i An in line ion chromatograph (IC) has been installed.
Functional testing of the IC has not been completed.
The valve repair will be performed upon receipt of parts.
Discussions were held with the staff to determine if the PASS capabilities were still consistent with the parameters described in Figure 1 of Inspection Report 50-312/87-05 and the effect of the design modifications.
The inspector was informed that the PASS capabilities described in Figure 1 of Inspection Report 50-312/87-05 were not correct.
The inspector informed the staff that Figure 1 was provided to the NRC inspector on February 9, 1987.
The staff stated that SMUD memorandum, RJR-86-97, submitted to NRR on March 13, 1986 identified that backup calculational methods for estimating the radioisotope concentrations in the reactor building atmosphere and the total dissolved gases in the reactor coolant were being deleted.
The staff apologized for providing the inspector with the wrong information in February 1987.
The inspector was also informed of SMUD memorandum, JEW 87-750, that was submitted to NRR on May 4, 1987.
The memo provides information on the accuracy and ranges of PASS that are not consistent with R.G.
1.97 requirements and that differ from the Districts previous letters of June 17, 1983 and March 13, 1986.
The three analyses
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most affected are boron, chloride and total gas.
The May 4, 1987 j
letter is currently being reviewed by NRR.
The inspector discussed the status of system testing.
In particular, the delays in resolving problems identified from testing of the containment atmosphere would appear to be unacceptable during normal plant operations.
The inspector observed that the status of procedure development, time and motion studies, decay heat sample testing, reactor coolant sample testing, and training of the PASS l
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operating staff appears to indicate that, unless the rate of progress increases, the system may not be ready to support the plant startup schedule of January 1988.
An inspection of the PASS system was also conducted.
The following observations were taade:
The sample point for obtaining a grab PASS sample was cluttered and is located in a contaminated area.
It appears that i
I unnecessary delays would be encountered should the need arise to obtain a sample.
This condition was identified during a previous. inspection.
There appears to be sufficient room in the area to provide clear access to the grab sample station.
l Temporary !ead blankets used to shield a portion of the sample skid are not installed in a permanent manner.
The blankets can be easily removed by anyone needing them for other purposes.
The inspector expressed concerns about this condition during previous inspections.
The location of the nitrogen supply appears to be in a location that would not be accessible to personnel during an accident condition ~due to the expected radiation levels.
Failure of the supply system or depletion of the nitrogen supply could cause delays under emergency conditions.
The above observations were brought to the licensee's attention.
The inspector was informed that at a recent SMUD management meeting the status of PASS testing had been discussed.
The licensee added that they are in the process of taking actions to closely monitor j
the status of PASS and to ensure sufficient resources and material
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are made available to complete the PASS testing in time to support plant startup.
This 4tcic will be examined prior to restart (86-37-01).
D).
Part 21 Report (Closed) 87-02-P1-A GA Technologies, Inc. report made pursuant to 10 CFR Part 21.21, identified a defect in the post-LOCA in containment high range monitoring system. The defective part was identified as the ion chamber's signal coaxial cable.
A licensee evaluation report, dated May 21, 1987, was reviewed.
The report include recent test data performed by the coaxial cable supplier that indicated that the current coaxial cable will meet R.G. 1.97 requirements throughout a postulated LOCA and are therefore acceptable.
This matter is closed (87-02-P1).
E).
Licensee Event Reports (LER)
(Closed) 86-19-L1 - LER 86-19 reported that grab samples were not taken, as required by T.S. Table 3.16-1, after the auxiliary building gas monitor R15002B was declared inoperable.
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The licensee's actions, which includes the replacement of monitor R15002, were consistent with the licensee's commitments described in the LER.
This matter is closed (86-19-L1).
(Closed) 86-20-Y0 - LER 86-20 identified inconsistencies with core damage assessments using methods developed by Stone and Webster.
This matter is being evaluated by NRR.
Regional action is closed (86-20-YO).
(Closed) LER 86-27-L0 LER 86-27, Revision 1 identified that reactor building duct particulate air sample filter data was lost.
The inspector verified that the corrective actions identified in the LER were completed.
The corrective actions were found to be satisfactory.
This matter is closed (86-27-LO).
12) Exit Interview The inspector met with licensee representatives (denoted in paragraph 1)
at the conclusion of the inspection on June 25, 1987.
The scope and findings of the inspection were summarized.
The licensee was informed of the violation discussed in section 6.
The inspector emphasized the importance for closely monitoring the status of PASS testing for the purpose of avoiding an unnecessary delay.
The inspector reiterated the concerns expressed in the cover letter of Inspection Report 50-311/86-11 for assuring the long term improvement discussed in section 11 are completed in a timely manner.
The licensee acknowledged the violation stating appropriate corrective action would be taken to resolve the observation discussed in section 6.
The inspector commended the Radiation Protection Group and Principal Engineering Technician for maintaining their responsibilities in the ALARA Program, l
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