IR 05000312/1981016
| ML20010B911 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 07/28/1981 |
| From: | Book H, Cillis M, North H, Wenslawski F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20010B908 | List: |
| References | |
| 50-312-81-16, NUDOCS 8108180471 | |
| Download: ML20010B911 (10) | |
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U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION V
Report No.
50-312/81-16 Docket No.
50-312 DPR-54 License No.
Safeguards Group Licensee:
Sacramento Municipal Utility District P. O. Box 15830 Sacramento, California 95813 Facility Name:
Rancho Seco Inspection at:
Clay Station, California Inspection conducted:
May 21-22 and June 2, 1981 Inspectors:
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M. Cillis, Radiation $peciali:t Date Signed l
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H. S. North, Radiation Specialist
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D6te Signed F/P/
Approved by:
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- F. A. Wenslawski, Chief, Reactor Radiation Protection Ifate Signed Secti,on Approved by-
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H. E. Book, Chief, Radiological Safety Branch
' Date' Signed Summary:
Inspection on May 21-22 and June 2,1981 (Report No. 50-312/81-16)
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Areas Inspected:
Reactive unannounced inspection by regional' based inspectors of the radiation protection and waste management aspects of a steam generator tube failure including:
steam generator tube failure, liquid and gaseous waste releases, independent measurement sample collection, training, steam generator repair activities, surveys, personnel monitoring, and facility tour.
The inspection involved 39 inspector hours on site by two NRC inspactors.
Results: Of the nine areas inspected, no items of noncompliance or deviations were identified.
8108180471 810730 PDR ADOCK 05000312 G
PDR RV Form 219.(2)
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DJTAILS 1.
Licensee Pe_rsonnel Contacted P. Cubre', Plant Superintendent
- R. Coluubo, Technical Assistant, Nuclear Operations
- R. Miller, Chemical and Radiation Suparvisor (CRS)
- F. Kellie, Assistant Chemical and Raolation Supervisor (ACRS)
M. Bua, Senior Chemical Radiation Assistant (SCRA)
D. Gardiner, SCRA T. Morrill, SCRA
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Newey, SCRA J. Bowser, Chemistry Radiation Assistant (CRA)
R. Bowser, CRA S. Manofsky, CRA R. Terpstra, CRA D. Stretars, Senior Typist Clerk D. Bird, ALARA Engineer
- D. Blacnly, Operations Supervisor M. Hieronimus, Shift Supervisor
- T. Perry, QA Supervisor
- H. Heckert, Nuclear Engineering Technician T. Grew, Welder (* Denotes those present at the exit interview.)
Non Licensee Personne'l Contacted H. Gooch, B&W, Senior Engineer
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2.
S_ team Generator Tube Failure
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On May 16, 1981, the reactor was 'at 98 percent power following the recent refueling outage -(Reference IE Inspection Report No. 50-312/
81-13). At 2200, the condenser air ejector (CAE) monitor reached the alert set point of 1000 cps.
The SCRA, telephoned by the operations staff at 2211, reported to the site within 15-20' minutes.
A high alarm was received on the CAE monitor at 2225.
At 2230 the operations staff stopped filling the condensate storage tank (CST)
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from the condenser hot well and started direct filling the CST from the raw water supply using mixed bed demineralizers.
This action, not part of the procedures, served to protect the contents of the CST from contamination. At 2235 the "B" steam line radiation monitor indicated ils mr/hr.
"B" main steam line was isolated at 2240. The staff began a controlled shut down at 2305.
The licensee notified the IE Headquarters duty officer of the occurrence at 2340. The reactor was tripped at 0055, May 17, 1981
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I and was at cold shutdown at 1005. When the tube failure occurred, eight of the nine condensate demineralizers (polishers) were in service. As the plant power level was reduced, the number of polishers in service was reduced to three; however, all polishers but one were contaminated as a result of the occurrence. At the
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end of the occurrence, both "A" and "B"l steam lines were filled with water to the turbine stop valves. Between May 17 and May 18, considerable additional' contamination of the secondary system with primary coolant occurred, as indicated by changes in once through steam generator (OTSG) boron concentration.
Secondary bide Boron Concentration (ppm)
OTSG "A" 0TSG "B" May 17 0'1700-
1250 May 18 @ 1900 900'
1250 This occurred in spite of the fact that
"B". steam generator was isolated at 2240 on May 16. At the time of the inspection, the licensee believed that-the additional contamination of "A" steam generator may have resulted 'from flow through the upper annulus drain valves of both steam-generators.. 'It was noted in the draft minutes of the May 27, 1981, PRC,! Meeting No. 928, that the PRC recomended that Operating Procedure No. -4.6 be revised to delete the opening of the annulus ~ drain valves.
No items of noncompliance were: identified. >
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3.
Liquid Waste Releases ~'
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The licensee established three committees to-respond to the principle problems presented by the occurrence;.a' Water 'Comittee, Plant Schedule Committee and Report Comittee.
In addition, surveys of the secondary side of the plant were perfonned to identify radiation
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levels on various components, crud traps, and leaking components.
Barriers were established to control access to components exhibiting significant radiation levels and areas where leakage could provide a source of contamination.
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The Water Comittee using component and system survey and sample analysis data established a plan for handling the lege volume of contaminated water. The plan incorporated three sequenced phases.
Phase 1 involved: draining the condenser hot well and feedwater systems through a series of sumps to the "A" or_"B" regenerant holdup tanks'(RHUT); dilution, mixing and sample analysis; discharge'to one of the two 500 k gal. retention basins;- dilution, mixing and sample analysis; and controlled discharge to the plant effluent stream which was maintained at approximtely 8500-9000 gpm with dilution flow from the Folsom South cenal. For Phases 2 and 3, the final processing of liquids released from the plant were the same from the point at which water was transferred to the "A" or "B" i-
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RHUT.
Phase 2 involved draining the secondary side of the!0TSG
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through demineralizers and-processing through the boric acid evaporator.
The condensate was discharged.to the "A" or "B" RHUT. Phase 3
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involved draining the main steam lines, processing through the
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the "A" or "B".RHUT.
The processing and discharge included not l
only the original contents but the water resulting from fill-flush cleaning activities.
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l Actions with respect to di'4ution and discharge were based on an ysis of samples collected from Jiluted, mixed tank or basin contents.
j Grab samples of the final diluted plant ~ effluent were collected
just prior to the effluents leaving the plant site. Calculations i
of total activity released frun the site were based on the analysis of the diluted RHUT tank contents.and the fraction of the tank
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contents discharged to the retention basin.
Liquid releases were performed in ac..:ordance with Plant Procedure AP 305-13, Environmental
Releases of Liquid Radioactivity.
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Analyses performed included tritium, gross beta, and gama spectrum.
Records of analytical results, volumes of tanks or basins discharged,
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release concentrations and dilutions were maintained. A cumulative sumary of individual basin releases and the fraction of the. Technical
Specification limit was maintained.. Technical Specification,
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Appendix B, 2.6.1, Specifications for Liquid Waste Discharges, limits discharges to; paragraph A, concentrations "not to exceed t
the values specified in 10 CFR 20, Appendix = B, Table II, Column 2",
and paragraph B, which limits releases of radioactive materials,
excluding tritium and dissolved gases, to 10 Ci/ reactor / calendar quarter. Liquid waste discharge records examined (LWR 81-1 May 18, 1981 i-through LWR 81-14, June 1, 1981) reported effluent concentrations less than the applicable 10 CFR 20, Appendix B values.
In addition,
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fission and activation product releases totaled 0.24 Ci or approximately
19.2 percent of 1.25 Ci.per calendar quarter, one fourth of the i
annual design objective of 5 C1.
No items of noncompliance were identified.
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Gaseous Releases l
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l The initial identification of the OTSG tube failure was' signaled by.
the CAE monitor.
In order to quantify the gaseous releases resulting from the occurrence, the licensee analyzed sample data and recorder -
i charts for the CAE, gland seal exhaust (GSE) and'the auxiliary
. building vent monitors. The CAE was sampled three times during the
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first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />..The nuclide mix, identified in the first sample taken at 2324 on May If as used to establish the monitor response i
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-4-rather than relying on the CAE monitor t *.libration based on equivalent Xe-133 concentration.
The licensee found that the CAE monitor (R-15004), a gamma detector, over responded by a factor of 5 to the mixture of nuclides released when compared with an equivalent Xe-133 concentration.
The gases released.via the CAE and GSE were vented through the auxiliary building vent where they wereragain monitored by the auxiliary building vent monitors.
The licensee' based the estimate of the quantity of noble gases released on the corrected CAE and GSE monitor results alone. The release measured by the auxiliary' building vent monitor was not corrected.
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The licensee believes that the' auxiliary building ve'nt monitor (R-15002',),
a beta sensitive scintillation detector, over' responded to the mixture released due to the high~ energy betas produced by some of the ntclides preseat.
The releases via the airborne pathway were substantially terminated with the loss of. condenser vacuum approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after the tube failure.
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The releases during.the 15 hour1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> period, calculated by'the licensee, were:
Curies CAE 4.12 GSE 0.29 i
Auxiliary Bldg. Vent 7. 5 The instantaneous release rate for. noble gases (Technical Specification, Appendix B, 2.6.3 A.) was less than one percent based on the CAE discharge alone.
No items of noncompliance were identified.
5.
Independent Measurements On June 2,1981, samples were collected from the plant effluent discharge during release of a retention basin (LWR 81-14).
Results of the comparison will be reported in a subsequent report (50-312/81-16-01).
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Training The inspector reviewed the training program implemented for steam-generator repair activities conducted by the licensee and contract
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personnel. The inspector reviewed training records and observed l
mockup training in progress at the time of the inspection. The f
training program for steam generator workers consisted of the i
following:
(1) Security training (2)
10 CFR 19.12 training
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(3) Respiratory training pursuant to 10 CFR 20.103.(c)
i (4) Mockup trainina for steam generator tuba pluggers (5) On-the-job instructions The training program for items (1) and (2) includes eight hours of classroom instructions and an examination and documentation of the
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i training effectiveness. The training program for item (3) includes
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two hours of classroom instruction, and examination and actual
wearing and fit-test of the various types of respirators used by
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the licensee. A review of the licensee's training records for items (1) through (3), lesson outlines and quizzes, indicated that all involved personnel had satisfactorily completed the required
training. The training provided was consistent with the requirements
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and recommendations of Sections 5.3.4 and 5.4 of ANSI-N18.1-1971.
Mockup training was jointly conducted by the licensee and Babcock
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and Wilcox (B&W) contract personnel to assure that ALARA criteria of 10 CFR 20.1(b) and R.G.-8.8 is achieved during steam generator-
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repairs. The mockup training for steam generator tube pluggers was observed by the inspector. The newly appointed plant ALARA Engineer also observed the training. The mockup training corsisted of:
(1) Personal Instructions by plant radiation protection and B&W personnel.
Detailed radiological control requirements and past-tube l
plugging experiences for accomplishing the steam generator repairs were discussed.
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(2) Dressing and undressing rehearsals using; protective clothing consisting of: two pairs-of plastic booties, rubber shoe
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covers, one pair of cloth, and.one pair of "tyvek,"' coveralls,.
cloth hood, air fed full-face respirator, or.e pair of-cloth gloves, four pair of rubber gloves,' and one pi'ece plastic wet suit with an attached hood." Dressing and. undressing,0f~ personnel
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(3)
Familiarization with tube plugging equipment'
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I (4) Welding time trials.
(The welding time trials of tube stabilizer I
j assemblies were performed on a bench top). The'best welding trial was 45 seconds.
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(5) Time trials for entering the steam generators and. installing (
i the 109 inch stabilizer assembly was conducted using the
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protective clothing described in paragraph (2) above. Special i
emphasis was placed on team work and the importance for exiting the OTSG should the worker ' encounter any problems. Time trials observed by the inspector. ranged from 45 seconds to
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120 seconds.
I The mockup training obser..
>peared to be meaningful in assuring that the ALARA crite ia is achieved.
No items of noncompliance were observed.
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7.
Tube Plugging'of "B" Steam Generator i
Steam generator repair operations in progress at the time of the.
inspection were reviewed by the inspector. The repairs were being performed jointly by the licensee and B&W contractor
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personnel. The inspector reviewed steam generator repair procedures
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i and observed some of the repair activities in progress at the time l
of the inspection.
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Procedure M.13, Rev. 4, "Once-through Steam Generator Tube ~ Plugging
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Procedure", was in use and provided detailed instructions addressing -
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the scope, precautions, special-tools and equipment, testing and acceptance criteria, leak testing requirements, manway cover removal / installation instructions, eddy current testing, installation and welding of
stabilizer rod assemblies, explosive plugging instructions, Quality
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Assurance requirements, and radiological control requirements.
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The plant radiation protection staff ~has coordinated closely with
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the planning and scheduling staff and Babcock and Wilcox consultant.
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Results of the meetings were documented and then discussed with
CRA's assigned to support OTSG repairs.
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Five (5) Radiological Work Permits for key 0TSG repair evolutions were reviewed by the inspector. The RWP's appeared to specify
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adequate radiological controls for accomplishing the repairs. All repair activities are being accomplished under the direct observation i
of plant radiation protection personnel. Manning by radiation i
protection personnel appeared to be adequate. The licensee had i
augmented the radiation protection staff with 5 senior and 3 junior i
contract radiation protection technicians. The contract technicians
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were retained by the licensee from the refueling outage which was l
recently completed.
All repair activities are being conducted in, "herculite," tent enclosures installed around the OTSG upper and lower manway.
The tents are equipped to accomplish the leak and eddy current
j testing and tube plugging repairs. An engineered ventilation
system pursuant to 20.103(b) equipped with a HEPA filter was installed on the tents. The tents also provided capabilities for obtaining
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i air samples of the work environment. Three individuals will be inside the tent and one individual will be outside the tent during i
entries inside the OTSG bowls. The four individuals will consist of two steam generator workers (one worker.will be located
inside the tent and provide assistance to the worker inside the
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0TSG bowl) and two radiation protection personnel (one inside
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the tent and one outside the tent).
j A leak test of the OTSG tubes was conducted and completed during
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The leak test identified one (1) leaking
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this inspection.
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tube which required plugging. Eddy current testing commenced upon completion of the leak test. The NRC inspector observed eddy current testing operations in progress during a tcur of the containment building. The eddy current evolution identified three (3) additional tubes which required plugging. Eddy. current operations were conducted by B&W while under the direct surveillance of licensee radiation protection personnel.
The four tubes were plugged several days after this inspection.
The tubes were plugged by licensee mechanics with the assistance of B&W personnel. The highest exposure received by an individual during this evoluation was 510 mrem based on pocket ionization chamber (PIC) a'nd thermoluminescent dosimeter (TLD) results. The licensee was currently in the process of closing up the steam generators during the inspection.
Two items of concern were identified by the inspector during the inspection.
One item required immediate consideration. These items were discussed with the licensee prior to the commencement of tube plugging operations.
(1) Comunication between the worker inside the OTSG bowl and ~the individuals inside the tent was inadequate. Verbal communications were impaired by the use of air supplied respirators.. Consequently a system had been developed to signal the worker by tugging on his breathing air line. This appeared unacceptable in the view of the possible separation ofthe individual's air supply line. The licensee agreed stating that a separate line would be attached to the individual.
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(2) The second item of concern' for future considaration involved the proper sequencing of tent installation,such that the ALARA considerations of R.G. 8.8 are implemented. -It appeared that the upper tent could have been in' stalled with the cover in. place. =The cover provides additional shielding, reducing the exposure received by the tent installers.
The licensee agreed stating that future procedures would be strengthened.
The licensee stated'at the exit interview that although the manway had been removed prior to tent installation, it h~ad been.epositioned to provide shielding, prior to tent installation.
No items of noncompliance were identified.
8.
Surveys The inspector reviewed radiation, contamination, and airborne survey records taken by the licensee in the primary side of upper and lower OTSG bowls and work tents. The highest penetrating (gamma) radiation levels detected were observed in the lower OTSG bowl. Contact readings
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taken on the tube sheet indicated levels of 14 rem /hr. General area readings taken in the center of the bcwl indicated levels of 10 rem /hr.
Levels of 4 rem /hr were detected at the manway opening and general area readings in the work tent ranged from 50 to 900 mrem /hr. The upper steam generator penetrating radiation levels are approximately 20%
less than the lower steam generator levels. TLDs were also exposed inside the steam generators to confirm the instrument measured radiation levels prior to allowing personnel entry. TLD results were in close agreement with the portable instrument measured radiation levels.
The highest non-penetrating (Beta) radiation levels were detected in the upper generator. Levels of 20 rad /hr were detected on contact with the upper tube sheet in comparison to 4 rad /hr detected on contact with the lower tube sheet. Loose surface contamination
levels were approximately the same. Levels ranged from 20,000 dpm/100 cm to 40 mrem /hr/ swipe.
Airborne concentrations inside the steam generator bowls were on the order of 10 times MPC for the most restrictive radionuclides without the use of respiratory equipment; however, personnel entering the work tents and steam generator were required to wear air supplied full face respiratory protective equipment which provided a protection factor of 2000.
No items of noncompliance were identified.
9.
Personnel Monitoring Personnel monitoring devices for steam. generator workers consisted of film badges, TLD's and PIC placed on the individual's hands, feet, thighs, chest, and head. The personnel monitoring prog.am was fiscussed
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in IE Inspection Report No. 50-312/81-02. The film badge results provide the individual's' official exposure. TLD and PICS are used to track an individual's exposure during the performance of critical operations and may be used to record the individual's exposure if the film badge is lost.
Prior to issuance of dosimetry and respiratory devices, new or contractor employees must have completed the radiation protection training, respiratory training, and have been counted in the on-site whole body counter. Whole body counting is required for all terminating personnel.
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The inspector confirmed that steam generator workers had completed the required training and reviewed their personal dosimetry records.
The personal dosimetry records reviewed appeared to be consistent with 10 CFR,20.102 requirements.
It was noted that the whole body counting required by the licensee's procedure #AP 305-20, "Whole Body Counting", was not performed for a terminating employee who had left the plant in April 1981. The employee has since returned to the plarm. Whole body counting for terminating employees may be missed for up to approximately 10% of terminating temporary employees due to the failure of temporary employees to inform the plant staff that they were terminating. This problem was discussed at the exit interview.
No items of noncompliance were identified.
10. Facility Tour The inspectors conducted a tour of the containment and turbine buildings making independent measurements and observations to determine compliance with the following regulatory requirements:
Topic Area Examined Requirement Posting of radiation areas, high 10CFR20.203(b),(c),(d),(e)
radiation areas, airborne activity, controlled a.'eas, and radioactive materials areas.
Control of radiation and high 10 CFR 20.105(b) (1) and (2)
radiation areas
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Engineered Controls
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10 CFR 20.103(b) (1) and (2)
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The independent measurements were made with an Eberline Model E-520 G-M survey meter, NRC serial number 006385, due for calibration on
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Independent radiation ' survey measurements made by the inspectors substantially
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confirmed the licensee's survey results and substantiated the licensee's posting and labeling practices. The inspector observed eddy current operations during the tour.
No items of noncompliance were identified.
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11.
Exit Interview At the conclusion of the inspection on May 22', 1981, the inspection
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findings were summarized for the individuals denoted in' paragraph 1.
The licensee was informed that some questions had been raised concerning the volume of water being discharged from the site.
Technical Specification, Appendix B, 2.0 Environmental Protection, section 2.1 Maximum Discharge Temperature, makes reference to a maximum discharge flow of 4000 gpm in the basis. However, no specific limit with respect to discharge from the site is associated with releases of radioactive effluents. The inspectors noted that the action which prevented contamination of the condensate storage tank significantly limited the quantity of contam '=+ed water requiring discharge or treatment. The inspectors informed the licensee that improved communications with steam generator repair personnel, to preclude tugging on the respirator air supply line, would be desirable. The license's attention was directed to the failure to capture all personnel,' principally contract workers, for termination whole body counts. The licensee stated that methods to improve communications with steam generator workers and capture personnel for whole body counts would be examined.
The results of the inspt: tion continuation on June 2,1981, consisting of a review of additto ei waste discharge records, and the collection of a discharge effluent sample, was discussed with the licensee by telephone.
No items of noncompliance were noted during that portion of the inspection.
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