IR 05000397/1986014
| ML20205C636 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 08/05/1986 |
| From: | Yuhas G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Gladhill R NATIONAL INSTITUTE OF STANDARDS & TECHNOLOGY (FORMERL |
| References | |
| NUDOCS 8608120419 | |
| Download: ML20205C636 (1) | |
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AUG 051986 National Bureau of Ftsndards Admin. A531 Gaither:; burg, MD 20899 Attention:
Mr. Robert Gladhill Gentlemen:
This refers to our telephone conversation of July 29, 1986. An enclosure to this letter contains a copy of Inspection Report No. 50-397/86-14 which you
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requested.
If yoh have additional questions, or if I can be of further assistance in this matter, please do not hesitate to contact me.
Sincerely, C' ?: ',' N,'.', D ny
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- C:: i~f i Greg Yuhas, Chief Facilities Radiological Protection Section
Enclosure:
As Stated
REGION V==
L GI' D Report No.:
50-397/86-14 p;'T* '
DESIGUMk0 hbh' !
, h.p" Docket No.:
50-397 Cert m ed M ;,
License No.:
NPF-21 Licensee:
Washington Public Power Supply System P. O. Box 968 Richland, Washington 99352 Facility Name: Washington Nuclear Project No. 2 (WNP-2)
Inspection at: WNP-2 Site, Benton County, Washington Inspection Conducted: April 28 - May 2, 1986 and Telephone Calls of May 8, 9, 12, 13, and 14, 1986 Inspectors:
k-f 30 f/e M. C'illis, Radiation Specialist Da'te Signed l'- 30 0 J. y Rus' sell, Radiation Specialist Date Signed Approved by:
@ k,h ff/30/M G. P.
s, Chief Da'te Signed Facili '
Radiological Protection Section
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Summary:
Inspection on April 28 - May 2, 1986 and telephone calls of May 8, 9, 12, 13, and 14, 1986 (Report No. 50-397/86-14)
Areas Inspected: Routine unannounced inspection by regionally based inspectors of radiation protection activities during refueling outage conditions; ALARA; external occupational exposures, licensee's action taken on followup items, licensee's action taken on IE Information Notices (ins);
control of radioactive materials and contamination, and a tour of the licensee's facility.
Inspection procedures 83724, 83726, 83728, 83729, 90713, and 92701 were performed.
Results:
In the six areas examined, two unresolved items were identified in one area: neutron dosimetry anomalies (paragraph 3); multi-whole body and extremity monitoring anomalies (paragraph 3); additionally, four open items associated with contamination control practices were identified (paragraph 5);
no violations were identified in the remaining areas.
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i DETAILS
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Persons Contacted
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a.
WNP-2 Staff
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- C. Powers, Plant Manager
- J. Baker, Assistant Plant Manager
- R. Graybeal, Health Physics / Chemistry Manager
- D. Larson, Manager of Radiological Programs and Instrument Calibrations
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C. Graybeal, Dosimetry Specialist
- G. Oldfield, Principal Health Physicist
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T. Chapman, Senior Health Physicist
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D. Ottley, Rad Services Supervisor
- L. Berry, Health Physics Supervisor
- L. Bradford, Assistant Health Physics Supervisor
- V. Shockley, Health Physics / Radiochemistry Support M. Valdez, Health Physics Foreman
- J. Mills, ALARA Coordinator D. Rinehart, Training Supervisor A. Hosler, Nuclear Safety Assurance Group Manager K. A. Smith, Health Physicist D. M. Truman, Health Physics Technician D. S. Feldman, Plant QA Manager M. Detrick, Health Physics Technician R. Wardlow, Health Physicist b.
Contractors
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(1) Allied Nuclear, Inc.
D. E. Krieselmeyer, Senior Radiation Protection Technician (2) Applied Radiological Controls, Inc.
L. A. Prictchard, Senior Radiation Protection Technician c.
U. S. Nuclear Regulatory Commission l
- R. T. Dodds, Senior Resident Inspector l
- R. C. Barr, Resident Inspector d.
State of Washington, Energy Facilities Site Evaluation Council
- W. Finch, Executive Secretary
- Denotes those present at the exit interview on May 2, 1986. '
In addition, the inspectors met with other members of the licensee's and contractor's staff.
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2.
Refueling Outage
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a.
Organization -
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The licensee's Health Physics (HP) ' organization established for the refueling outage has not changed from that described in paragraph 2(b) of Inspection Report 50-397/86-10. -
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No violations or deviations were identified.
b.
Refueling Work Schedule The inspectors were informed that the HP staff was working twelve hour days, six days per week during the outage.
An examination of personnel time keeping records disclosed that five individuals worked in excess of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per week for the pay period ending April 6, 1986, and two individuals worked in excess of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> per week for the pay period ending April 20, 1986.
The inspectors verified that authorizations for the overtime work hours were consistent with plant procedure 1.3.27, " Overtime
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Control" and Technical Specifications, Section 6.2.2(f).
No violations or deviations were identified.
c.
Staffing of Contract HP Technicians Discussions with the licensee's staff disclosed that the plant's l
permanent staff of 16 Health Physics Technicians (HPT) was augmented with the addition of 32 Senior and 11 Junior HPIs from three
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different sources. Additionally, 11 Plant Chemistry Technicians I
were temporarily assigned to the Health Physics group during the j
outage.
l The resumes of 43 contract Senior and Junior HPTs were reviewed.
Additionally, the inspectors met with some of the contractor HPT staff during a tour of the licensee's facility.
The review found that all Senior HPT net the qualifications for technicians in a responsible position as recommended by paragraph j
4.5.2, American National Standards Institute (ANSI)/ANS 3.1-1978,
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" Selection and Training of Nuclear Power Plant Personnel."
On two occasions during the inspection, the inspectors noted that the day shift staff of HPT were unable to support all work requests for HP coverage. The work was subsequently supported when some HPT became available.
No violations or deviations were identified.
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d.
Personnel Exposure Radiation exposure records of the licensee's and contractor's staff
were reviewed.
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The year-to-date exposures were well below the limits prescribed in i
10 CFR Part 20.101, " Radiation Dose Standards for Individuals in j
Restricted Areas."
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No violations or deviations were identified.
e.
Personnel Contaminations Personnel contamination records for 1986 were reviewed.
The review disclosed that a total of eighteen skin contamination occurrences were reported. Each of the occurrences was investigated
.in accordance with procedure PPM 11.2.13.3, " Personnel Contamination Survey."
The maximum skin contamination identified to date was reported as 4E4 dpm/ probe area. All individuals were decontaminated with minimum effort.
The inspectors noted that on August 20, 1985, the ALARA Committee established an ALARA goal for maintaining skin contaminations that require dose assessments to one per month. The inspector informed the licensee staff that this goal secoed unnecessarily permissive.
The inspector also discussed this at the exit interview. The inspector stated it would be prudent to establish a more conservative goal.
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The inspectors noted that the licensee staff carefully review and
trend all personnel' contamination occurrences. Plant management is made aware of all occurrences.
No violations or deviations were identified.
f.
Outate Management Meetinas Outage management meetings are conducted two times per day. The status of the outage, scheduling, ALARA, and any other problem related to the outage are discussed at the meetings.
The NRC inspectors attended two management meetings during the inspection.
No violations or deviations were identified.
3.
External Exposure Control
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The inspectors examined the licensee's personnel monitoring program for control of external exposure during normal and outage operations for compliance with 10 CFR Part 20 requirements. Representative radiation
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exposure records, licensee's procedures and licensee's memoranda were reviewed. Discussions with cognizant licensee staff and management were held and their facilities for control, distr.ibution and processing dosimetry were toured. The examination, disclosed the following:
The licensee's exposure control program utilizes Teledyne Isotopes
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a.
CaSO :Dy - Teflon TLD badges for personnel beta / gamma, neutron, a /nd j
environmental gamma monitoring; LIF-Teflon discs for extremity l
monitoring and Pocket Ionization Chambers (PICS) of various ranges
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i for dose control between monthly TLD reading and reporting periods.
j The examination raised several questions relative to the licensee's i
neutron, extremity, and multiple badging exposure monitoring program.
(1) The doses as read from a majority of neutron albedo TLDs during a recent job were a factor of 2-3 different from estimated doses based on timekeeping and measured area neutron dose
rates.
In over 80% of these cases, the TLD dose was higher
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than the timekeeping dose and on the remainder, it was lower.
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TLD doses ranged from 64 to 875. ares..
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L (2) Extremity dose as measured by TLD was significantly lower and often zero after work in very high radiation areas where dose to the head, torso, and thighs as measured by personnel beta / gamma TLDs were from 100 to 400 mrem.
In those cases
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where the extremity dose was not zero, doses to the hands averaged a factor of approximately 2 lower than doses to the
head, torso or thighs. Also, unused extremity TLDs read to provide a " background" reading along with batches of used extremity TLDs gave unusually high background readings, up to 91 arem for a period of less than 1 month since annealing.
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(3) Doses read from PICS were frequently a factor of 2-3 times higher than head, torso and thigh TLDs and 5-10 times higher than extremity TLDs.
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b.
The licensee was aware of the problems with the neutron and multiple
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badging exposure monitoring and had been taking action to resolve these questions. Further dose rate measurements and neutron spectrum distribution characterization are planned the next time the plant goes up in power to try and resolve problem (1) above. A i
memorandum was issued on April 29, 1986, by the Manager of
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Radiological Programs and Instrument Calibration which resolved the-PIC and TLD discrepancies during multiple badging, problem (3)-
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above, as being due to compounding reader errors from multiple entries and use of 0-5 Roentgen PICS. The licensee was not aware of the unusually low extremity doses and high background readings until-it was pointed out by the inspectors. The licensee said he would take action to resolve these questions. All other aspects of the licensee's personal dosimetry program for external exposure Appeared to meet the regulatory requirements. There was no indication of
exposures in excess of 10 CFR Part 20 limits to personnel as a result of these problems.
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c.
The inspectors examined the licensee's administrative controls for external exposure control to assure they were designed to maintain exposures ALARA. Other than the problems noted in the ALARA section
of this repor't, the licensee's activities were found to meet the regulatory requirements.
d.
The inspectors extmined the licensee's records and reports of i
external exposures and found that they met the regulatory
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requirements.'
i The problem identified with neutron exposure monitoring is considered an unresolved item (86-14-01). The problem identified with extremity exposure monitoring is considered an unresolved item (86-14-02).'
4.
Internal Exposure Control and Control of Radioactive Material The inspectors examined the licensee's internal exposure control program for compliance with 10 CFR Part 20.103, " Exposures of Individuals....In Air in Restricted Areas." The licensee's program for controlling radioactive material and contamination control practices were also examined. Representative records, procedures and reports were reviewed.
Discussions were held with cognizant licensee staff and management and their facilities were toured. An inspector also attended the licensee's j
respiratory protection training course. The following observations were made:
a.
Respiratory Protection and Whole Body Counting Programs The inspectors toured the licensee's body counting and respiratory i
protective equipment cleaning, maintenance, and testing facilities.
The inspectors reviewed records of air sampling and whole-body
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counts. No exposures in excess of 40 MPC-hrs were found and none of the air samples in the plant during work were in excess of 10 CFR 20
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Appendix B limits. 'The inspectors reviewed records and RWPs, contacted staff and management personnel, and one inspector attended the licensee's Basic Respiratory Protection training. The use of respiratory protection equipment appeared to meet the regulatory requirements prescribed in 10 CFR Part 20 and were consistent with the recommendations provided in Regulatory Guide 8.15, " Acceptable Programs for Respiratory Protection" and NUREG 0041, " Manual of Respiratory Protection Against Airborne Radioactive Materials."
No violations or deviations were identified.
b.
Procedures The following procedures from the licensee's Plant Procedures Manual (PPM) and Health Physics Description Manual (HPD) that are related to respirator protection program were reviewed:
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HPD 3.1.11, " Respiratory Protection"
PPM 11.2.4.1, "MPC - Hour Log"
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l PPM 11.2.11.2, " Selection of Respiratory Protection Equipment,"
Revision 2
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PPM 11.2.11.3,. " Issuance of Re,spira' tory Protection Equipment"
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' PPM 11.2.11.4, "Use of Respiratory Equipment, Cleaning,
Inspection, and Maintenance"
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PPM 11.2.12.2, " Selection of Protective Clothing"
HPD 3.1.15, " Contamination Control"
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PPM 11.2.8.1, " Radiation Work Permit"
PPM 11.2.13.2, " Personnel Contamination Survey
PPM 11.2.13.1, " Direct Area Radiation and Contamination Surveys
i PPM 11.2.15.7, " Release of Material from Radiologically
Controlled Areas" The following observations were made:
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(1) Paragraphs A.I.C.2 and B.I.C.2 of Attachment B to procedure PPM 11.2.11.2 reference procedure number PPM 11.2.12.1.
This is incorrect.
It should be PPM 11.2.12.2.
l (2) The surface contamination levels used for prescribing air purifying respirators are provided in PPM 11.2.11.2, paragraph B.1 and in Table 1 of PPM 11.2.12.2.
The inspector noted that the values provided in these sections do not agree with one
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another.
(3) The licensee's' grab air sampling program prescribed in the PPM or HPD Manuals do not appear to adequately consider or be consistent with the guidelines provided in Section 4.0 of ANSI N13.1; " Guide to Sampling Airborne Radioactive Materials in i
Nuclear Facilities." In particular, Section 4.2,
" Representative Samples"; Section 4.3, " Sampling Programming";
Section 4.3.5, " Specific Nature of the Operation or Process";
and Section 6, " Validation of Sampling Effectiveness." For example:
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The PPM or HPD Manuals do not appear to provide adequate
guidelines for obtaining representative. air samples for operations that are or may be conducive to creating airborne concentrations of radioactivity.
The PPM or HPD Manuals do not appear to provide adequate
guidelines for increasing the frequency for samplihg during work operations that are conducive of creating
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frequency would better assure the representativeness of l
the sampling program.
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The PPM and HPD Manuals do not appear to provide adequate
guidelines for carefully evaluating the nature of operations which create the p,otential for airborne
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i The PPM and HPD Manuals do not appear to provide adequate
guidelines for validation of the effectiveness of the
respiratory protection equipment during abnormal or unplanned airborne releases.
J (4) The PPM and HPD Manuals do not include provisions for assuring that respiratory equipment prescribed for use on Radiation Work
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Permits (RWPs) was worn by the individuals performing the work
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authorized by the RWP. Currently, the licensee does not maintain records of respiratory equipment issue or use.
(5) The inspector identified that procedures HPD 3.2.15 and PPM 11.2.15.7 allow the release of material having 5000 disintegration per minute (dpm) of fixed beta-gamma l
contamination.
It should be noted that IE Information Notice l
85-92, " Surveys of Waste Before Disposal from Nuclear Reactor l
Facilities" recommends in practice that no radioactive material (licensed) means no detectable radioactive material.
The above observations were brought to the licensee's attention during the inspection. The inspector was informed that the observations would be evaluated. The Health Physics / Chemistry Manager informed the inspectors that procedures HPD 3.1.15 and PPM 11.2.15.7 would be changed to conform with IN 85-92 recommendations.
(6) Internal Exposure The inspectors reviewed the licensee's procedures for control of internal exposure. The inspectors noted that the licensee's criteria for use of filtered-air respirators is based on a dual limit, i.e., measured airborne concentrations in excess of 25%
of the Appendix B limit and contamination levels in excess of
15 mrad /hr or 50,000 dps/100 cm fixed or loose, on surfaces being worked or in the work area, respectively. However, the use of respirator protective equipment with a protection factor greater than 50, e.g., supplied air masks, is based on having measurable or expected airborne concentrations greater than 50 MPC. The inspectors brought to the licensee's attention that, as high airborne radioactivity levels may occur due to worker's activity during the job rather than always being detectable by air sample prior to work, a similar contamination limit for use of supplied air systems would be prudent. Administrative control of internal radiation exposure appeared to meet the regulatory requirements.
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(7) Audits
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Annual audits of the respiratory grotection program are performsd by the licensee's Quality Assurance group. An audit of the program was in progress'at the time of this inspection.
i Prior QA audits related to this topic identified that the licensee's program was essentially consistent with the regulatory requirements and with plant procedures.
5.
Abnormal Contamination Control Occurrences a.
April 16, 1986, Occurrence On April 17, 1986, the licensee's Health Physics (HP) staff notified Region V that a spread of radioactive contamination occurred in the
Reactor Building from operations performed in the 606' cavity on April 16, 1986. Contamination levels ranging from less than.1E3
disintegrations per minute per 100 square centimeters (dpa/100 cm )
2 were found at the 471', 501', 522', 548', 572',
to 3E4 dpm/100 cm and 606' levels of the Reactor Building. The contamination was discovered after several workers were found to have contamination on their shoes as they were attempting to exit the licensee's controlled areas. The licensee's HP staff attributed the occurrence to a~1 operation associated with the cleanup of the reactor cavity.
An examination of the event was conducted during this inspection.
The examination included:
Holding discussions with the licensee's staff
Review of licensee's procedures referenced in paragraph 4(b)
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above Review of survey records
Review of PPM 11.2.19.1, " Investigation of Non-Reportable Radiological Occurrences" and Health Physics Program Description (HPD) 3.1.19, " Radiological Investigations, Audits, and Inspection" Conducting a tour of the affected areas
Review of Radiation Work Permits 2-86-00137 and 2-86-00159 The inspectors verified the accuracy of the information provided to Region V on April 17, 1986. Additionally, the examination disclosed the following:
Two maintenance workers were assigned to clean the reactor
cavity on April 16, 1986. Continuous coverage of the cleaning was provided by a plant HP Technician who met the qualifications recommended by American National Standards institute (ANSI)/ANS 3.1-1978.
The maintenance workers wore protective anti-contamination clothing and air-purifying i
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respirators even though the RWP (2-85-00159) did not require they be worn. Known contamination levels in the cavity, prior
to cleaning, ranged from 2E3 dpm/1,00 car to 15 mrad /100 cm,
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A prejob briefing was he'ld with the HP Technician, workers, and
workers' supervisor in attendance. The scope of the job was discussed at the meeting. The HP Technician requested that a controlled vacuum cleaner be used to' clean the cavity. A vacuum was not immediately available so the HP Technician
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relented to the pressure of getting the job done in order to l
support defueling operations. The Technician authorized the use of a foxtail (i.e., broom) and dust pan for the cleaning.
The Technician thought he could adequately control the work because the cavity air sweep ventilation system was operational.
The use of other filtered portable 2,000 cubic feet per minute
(CFM) units was not implemented even though several units were available at the site. This decision was not consistent with procedure PPM 11.2.11.2 which states:
"All practicable engineering or process controls have been used to reduce or eliminate the respiratory hazard,..."
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The work was accomplished in approximately 40 minutes to one
hour. Only one ten minute grab air sample was taken at the
start of the cleaning evaluation. A portable continuous air monitoring (CAM) system was located on 606' floor level of the Reactor Building. No additional grab air samples were taken during the cleaning. The one grab air sample taken was measured as 6.92E-11 microcuries per milliliter (uci/al).
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At change of shift, several workers exiting the plant were found to have contamination on their shoes. No individual, including the involved workers and HP Technician, were found to
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have skin contamination.
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An investigation was made by the licensee's HP staff. The results of the licensee's investigation were transmitted to the NRC Region V staff on April 17, 1986. The investigation was still incomplete at the conclusion of this inspection.
- The CAM on the 606' floor had indicated an increase of approximately 1,000 counts per minute (cpm) at the time the cleaning took place. The licensee's staff surmised that an airborne release of unknown levels had occurred as a result of the cleaning.
It was also surmised that the one grab sample taken may'not have been representative of the entire cleaning operation.
- None of the workers were given whole body counts. The '
inspector discussed the above problems and the concerns with the PPM procedures identified in paragraph 4(b) above with the Health Physics / Chemistry Manager. The inspector asked the Health Physics / Chemistry Manager whether or not the HP
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Technician used good Health Physics practices under the circumstances associated with this particular occurrence. The
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Health Physics / Chemistry Manager s,tated that in retrospect, the
HP practices could have been improved upon.
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The two workers were signed in on two different RWPs. The actual RWP for the work (i.e., 2-86-00159) did not require continuous Health Physics surveillance or any respiratory equipment. The HP Supervisor stated that the HP Technician did provide continuous surveillance and that respiratory equipment
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was provided to the individuals. He added that the Technician did not have time or may have forgotten to modify the RWP.
The inspector was informed on May 9, 1986, that the individuals
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involved in the April 16, 1986, occurrence ~were given whole body counts of which the results did not reveal any internal uptake of radioactivity.
j The licensee's investigation included an inspection of the Reactor
Building ventilation system. The licensee's inspection of the
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ventilation system was still in progress at the time of this inspection.
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The inspector discussed some of recommendations provided in Section
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18, " Surface Contamination and Decontamination" and Section 22,
" Control of Radioactive Air Pollution" of Radiation Hygiene Handbook by Blatz, 1959 with the Health Physics / Chemistry Manager. The inspector also discussed:
(1) the prerequisite in procedure PPM 11.2.11.2 for using all practicable engineering or process controls to reduce or eliminate the respiratory hazard, and (2) of 10 CFR
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Part 20.103(b)(1) which is consistent with PPM 11.2.11.2.
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inspector reminded the licensee's staff of the plant procedures HPD 3.1.19 and PPM 11.2.19.1 which require investigations be performed and documented for any deviations from good health physics practices. HPD 3.1.19 identifies such events as " lapse of contamination control within the restricted area" as radiological occurrences that shall be investigated. The inspector then asked
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the licensee's staff if they felt the health physics practices related to the specific event were adequate for reducing the respiratory hazard. The inspector added that their answer should consider those individuals working in adjacent areas who were not
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equipped with respiratory equipment during the postulated release of airborne radioactivity. The licensee's HP staff responded by
stating the ongoing investigation of the event will include an investigation of the health physics practices implemented during the cleaning of the cavity. The licensee's staff added that the j
investigation will be made in accordance with procedures HPD 3.1.19 and PPM 11.2.19.1.
The inspector informed the licensee that this item will be examined during a subsequent inspection (86-14-04).
b.
May 11, 1986 Occurrence j
On May 12, 1986, the inspector was informed of another occurrence involving the spread of contamination to the same areas identified
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during the April 16, 1986, event.
Contamination levels spread to
adjacent areas ranged to approximately SE4 dpm/100 cm.,- The events were similar except for the evaluation being performed and the way
in which the problem was identified.
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The licensee's staff indicated that the cavity was being drained down on May 11, 1986. The cavity walls were hydrolased during the drain down. The spread of contamination was found during the performance of a routine contamination survey. Nine individuals were given whole-body counts. The results of the whole body counts did not reveal any uptake of radioactivity by any individual.
A preliminary licensee investigation conducted subsequent to the May 11, 1986, occurrence revealed that the Reactor Building ventilation system flow pattern and exhaust flow rates'were not consistent with the system design specifications. The cavities two 8,000 cubic feet per minute (CFM) ventilation exhaust units were running at 4,500 CFM. Additionally, it was reported that the air supply on the 606'
was estimated to be 53,000 CFM while the exhaust was measured at 51,000 CFM.
On May 12, the licensee also reported that a split seam was discovered in the Reactor Building ventilation system ducting on May 6, 1986. The split seam was located on the discharge leg at the 572' level of the Reactor Building. The seam was repaired prior to starting the cavity drain down operation on May 11, 1986. The licensee's staff ruled out the split seam as the source of contamination seen on April 16, 1986. The rationale used for
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arriving at this conclusion was the contamination spread seen on May 11, 1986 (i.e., after the seam was repaired) and contamination surveys of the seam did not indicate any contamination levels above
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background. The licensee's staff concluded that the most probable cause for the April 16 and May 11 events was a misaligned damper and the inconsistencies'seen in the flow rates and flow patterns. The licensee informed the inspector that their evaluation of the Reactor
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Building ventilation system was expected to continue. The inspector informed the licensee that the results of their evaluation would be examined during a subsequent inspection (86-14-03).
c.
April 18 - 22, 1986 Occurrence (1) General During a tour of the Radwaste Building on April 30, 1986, the inspectors noted that an area was being controlled for contamination purposes that was not controlled during a previous tour made during the week of March 24 - 28, 1986. The area under control was the waste solidification and waste storage area located on the 437' level. An examination was conducted to determine the reason for the change. The '
examination included:
Discussions with the licensee's staff
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Tour of the area
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Review of survey records
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Review of the ALARA revie'w, dated April 17, 1986, for
disassembly of the reactor recirculating pump internals Review of procedures identified in paragraph 4(b) above
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Review of RWP 2-86-00165 j
The examination disclosed the following:
The 437' level was inadvertently contaminated during the
decontamination, disassembly, and inspection of the l
recirculation pump internals which was performed over the weekend commencing April 18, 1986. The work was performed in the licensee's equipment decontamination facility located on the 467' level of the Radwaste Building. The decontamination area is located directly above the contaminated area that was noted by the inspectors during I
the April 30 tour.
The licensee's staff informed the inspectors that the disassembly and inspection of the pump was originally scheduled to be performed sometime after the refueling outage (i.e.,
July, August, etc.).
Subsequently, a decision was made to accomplish the inspection before the new internals were reinstalled. The staff stated that they had originally planned to accomplish the work in a tent with a controlled ventilation system. The staff added that they were unable to implement
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i their original plans because of the time constraints to accomplish the inspection. Therefore, the staff informed WNP-2 management that the entire decontamination facility would be i
controlled as a " tent" and that the work would be accomplished under continuous HP surveillance.
The staff indicated that the pump's manufacturer informed the licensee that it would not take more than fifteen minutes to remove the pump's impeller. Additionally, the HP staff required that the pump internals be decontaminated and be maintained wet during the inspection. No additional engineering controls or other process controls were prescribed.
The staff expected that the facility ventilation system to be adequate for the work.
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The pump was transferred from the drywell.to the 467'
decontamination facility on April 17, 1986. A cart was used for the transfer. The pump was transferred via an equipment hatch (i.e., door) opening located between the 437' level and 467' level. The seam along the equipment hatch door was taped
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after the internals were transferred.
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The decontamination, disassembly, and inspection of the pump l
l got started on Saturday, April 18, 1986. Personnel wore I
protective anti-contamination clot,hing with wet suits and air purifying respirators to accomplish the work. The
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licensee's staff indicated tha't the decontamination went well; however, they were unable to remove the impeller as was originally planned (i.e., approximately 15 minutes). The staff indicated that the impeller locking ~ unit (i.e., nut) was frozen, and it took approximately two days before it was removed. The nut was subsequently removed with an air impact wrench. The impeller was then removed from the housing with a mechanical jacking device. The HP staff stated that
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approximately one pint of radioactive liquid drained on the deck when the impeller was removed. The water was cleaned up.
On April 20, 1986, the licensee's. staff found contamination levels on the 437' level ranging up to approximately 1E4 dpm/100 ca. The contamination was found by a HPT performing a t
routine survey.
I An investigation accomplished by the licensee's staff concluded that the cause for the contamination on the 437' level was attributable to the pump work that.was performed directly above
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(i.e., 467' level). The licensee's staff stated that a seal was originally planned to be installed around the equipment
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hatch opening. They added that it was never installed.
Survey records revealed that there was wide spread of contamination throughout the decontamination facility.
Smearable contamination levels ranged up to 720 mrad /100 cm,
All of the swipes were taken on the deck. The licensee's staff
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did not have any data to indicate contamination levels on the walls or in the overhead on the deck below (i.e., 437' level).
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There were no indications of any personnel contamination occurrences.
l The highest airborne concentrations measured during the work
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were as follows:
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Measured Nuclide Concentration (uCi/ml)
MPC (uCi/al)
% MPC Manganese-54 1.52E-10 4.0E-8 0.4 Cobalt-58 6.29E-10 5.0E-8 1.3 Cobalt-60 2.39E-9
.9.0E-9 26.6 Zine-65 3.25E-9 6.0E-8 5.4 It should be noted that the ALARA review predicted that the man-Rem estimate for accomplishing the work was approximately 2 man-Rem. Approximately 15.0 man-Rem had been expended at the time of this inspection. The work was not yet completed. The Health Physics / Chemistry Manager stated that he informed the Plant Manager that no additional work may be accomplished until the engineering staff clearly defines the work scope in very
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i specific detail. The Plant Manager agreed with the recommendation made by the Health Physics / Chemistry Manager.
Thelicensee's.staffresponsibleIorplanningtheworkraised
some concerns about the' accura'cy of the information provided to them during the preplanning stages by their engineering staff.
They added that the scope of the work changed during the course of the job. This tended to complicate their eff6rts performed between April 18 - 22, 1986. The inspectors observed that decontamination efforts were still in progress during this inspection.
The inspectors questioned whether actions were planned for providing a positive seal for the equipment. hatch door used to j
transfer the pump from the 437' level'to the 467' level. The j
inspectors also questioned the adequacy of the planning (i.e.,
consistent with good health physics and ALARA practices) that
was provided prior to allowing the work to start. The concerns
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discussed in paragraph 5(a) related to the April 16, 1986, were reiterated to the licensee's staff and at the exit interview.
The inspector pointed out a simple method that could have been used in preventing the April 16, 1986, and this occurrence.
j The inspector added that the event resulted in unnecessary personnel exposures, costs and potential of contamination of personnel working in adjacent areas.
The inspectors emphasized the need for improving their t
contamination control program. The need for training personnel
in the implementation of good contamination control practices was also brought to the licensee's attention.
- The licensee's staff informed the inspectors their current investigation of this event was still in progress at the time of this inspection. They added that the investigation will be made in accordance with HPD 3.1.19 and PPM 11.2.19.1.
The inspector informed the licensee that this ites will be examined during a subsequent inspection (86-14-05).
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The inspectors were also informed that the licensee's investigation would include the actions taken with regard to the equipment hatch seal. The inspectors informed the licensee that this item would be examined during a subsequent inspection
(86-14-06).
6.
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a.
General
An examination of the licensee's program for maintaining exposures
"As Low as Reasonable Achievable" (ALARA) was conducted.
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The examination included the following:
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Discussions with the ALARA Coordinator
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Review of ALARA Committee Meeting Minutes Review of procedure PPH 1.1.6, "Plynt ALARA Committee"
Review of procedures:
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11.2.2.1, " Apparent Reduction Potential (ARP)"
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11.2.2.4, "ALARA Improvement Suggestion" 11.2.2.6, "ALARA Outage Planning" 11.2.2.7, "ALARA Procedure Analysis"
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11.2.2.9, " Estimation of Achievability Index" Review of the Final Safety Analysis Report (FSAR), Section
12.1, " Assuring that Occupational Radiation Exposures are i
ALARA" I
Review of RWPs
Review of procedures:
HPD 3.1.2, "As Low As is Reasonably Achievable (ALARA)"
HPD 3.1.20, " Radiological Training" The inspectors found it very difficult to clearly define the ALARA program because of the four or five documents involved.
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No violations or deviations were identified.
b.
Organization and Responsibilities
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Procedure HPD 3.1.2 assigns the Health Physics / Chemistry Manager the responsibility for designating a Plant ALARA Coordinator.
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Administrative procedure PPM 1.1.6, " Plant ALARA Committee" establishes the ALARA Committee and their responsibilities. Volume 11 of the PPM establishes the ALARA implementing procedures.
Procedure HPD 3.1.20, Section 7, establishes the ALARA Design Training program.
The inspectors noted that none of the above procedures clearly identify the ALARA organization or who is responsible for implementing the ALARA program. HPD 3.1.2 clearly defines the responsibilities for the Director, Support Services; Manager, Radiological Programs; Plant Manager, Engineering, and Program Directors; but fails to define the responsibilities of the Plant j
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ALARA Coordinator.
j The ALARA Coordinator was unable to clearly define his responsibilities to the inspectors during the inspection. However;
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the inspectors did note that the ALARA Coordinator did perform functions related to the implementation of the ALARA p{ogram.
The inspectors noted that bits and pieces of the ALARA Coordinator's responsibilities are defined in sev'eral of the implementing procedures and in Administrative Procedure PPM 1.1.6.
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The inspectors infe rmed the licensee it would be prudent to clearly define who is responsible for implementing the ALARA program and the Plant ALARA Coordinator's responsibilities.
The inspector noted that the ALARA organization' staffing appeared to be marginal. The ALARA organization primarily consists only of the Plant ALARA Coordinator. Additional support is provided to the ALARA Coordinator from the Health Physics / Chemistry Managers during peak work periods. Limited support is-also provided by the Plant ALARA Committee members.
No violations or deviations were identified.
c.
ALARA Goals Discussions with the Plant ALARA Coordinator revealed that a goal of 292 man-Rem had been established for the refueling outage work package. A total of approximately 91 man-Rem had been expended by May 2, 1986. The status of the man-Rem estimates are discussed daily at the morning and afternoon management meetings.
The bases for the 292 man-Rem estimate was examined. The inspectors noted that the bulk of the 292 man-Rea estimate is based on three jobs. The estimates for the three jobs totalled 222 man-Rem. One
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job had been completed with a savings of approximately 22 man-Rem, another job was expected to complete for a savings of approximately
90 man-Rem, and the third job had just started.
The inspectors noted that no adjustments were planned t'o be made to the 292 man-Rem t
estimate. The man-Rem estimates were high because it had been I
assumed that the outage would start immediately after plant shutdown. However, the outage did not start for approximately 2 - 3 l
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weeks which allowed the dose rates to be lower due to decay. The inspectors informed the licensee that it would be prudent to include provisions in ALARA implementing procedures to allow for in process checks when it appears that estimates may be too low or too high.
The inspectors added that making such adjustments to provide more
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realistic ALARA goals would give management a more accurate assessment of the ALARA program.
I No violations or deviations were identified.
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d.
Audits The inspector held discussions with the Plant Quality Assurance (QA)
Manager. The discussion was related to audits and surveillance of
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Plant ALARA program.
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No violations or deviations were identified.
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Training
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Discussions were held with the Plant ALARA Coordinator to ascertain if the ALARA Training Program was consistent with Section 3.1.2.3'of procedure HPD 3.1.2 and as described in procedure HPD 3.1.20.
The
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Plant ALARA Coordinator did not appear to have a good understanding
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of the ALARA Training Program described in these procedures.
Subsequent discussions with the Training Supervisor revealed that the ALARA Training Program was being effectively implemented. The inspectors noted that the level of training provided is commensurate with the individuals position and/or responsibility. The Plant ALARA Coordinator had attended the training on two different occasions.
i No violations or deviations were identified.
f.
ALARA Committee Meetings and ALARA Evaluations The inspector verified that ALARA Committee meetings were held at the frequency specified in PPM 1.1.6.
ALARA Committee meetings for l
the period 1984, 1985, and 1986 were reviewed.
l The ALARA Coordinator provided the inspector with copies of four ALARA reviews. Three of the reviews were for jobs estimated to be in excess of 10 man-Rem and the fourth was for a job estimated at
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1.2 man-Res. The ALARA Coordinator stated the exposures for the
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last job had reached approximately 15 man-Ren and the work was still incomplete. He added that the Plant Manager had authorized that work be discontinued until the scope of the remaining work has been clearly identified. The reason for this is to assure an appropriate ALARA review could be performed in accordance with the ALARA concept. The inspector asked the Plant ALARA Coordinator if an
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ALARA Committee meeting was conducted for all 1986 scheduled work l
estimated to be in excess of 10 Man-Res. Although the reviews were conducted, the ALARA Coordinator originally informed the. inspector that he thought the cammittee failed to evaluate the three jobs.
The inspector noted that the ALARA Coordinator chaired the meeting that evaluated the three jobs. The inspector noted that the minutes i
failed to provide any specific details. The same minutes appeared to be cryptic. They did not include the man-Rea estimates or the
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disposition of whether or not the ALARA Committee members approved the estimates.
It appears that ALARA Committee meetings minutes are not consistent with Section F of paragraph 1.1.6.3 to procedure PPM 1.1.6, which states that the minutes should include:
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" Subjects considered and the disposition of each."
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The inspector noted that the ALARA implementing procedures currently
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do not provide the mechanism for performing in process; reviews during the performance of work when it yppears that the exposure
estimates for the job are in. jeopardy. The inspector informed the
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licensee that it may be wise to per' form such in process reviews for the purpose of regrouping and assuring there are no other ALARA considerations that need to be evaluated.
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The above observations were brought to the licensee's attention at the exit interview. The inspectors emphasized that the findings were strictly observations which were being brought to their
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attention for evaluation as areas that could improve the ALARA
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program.
No violations or deviations were idantified.
7.
Facility Tour The inspectors toured the Turbine, Radwaste, and Reactor Buildings, the Control Room, the Reactor Building Wet Well and the Dry Well and held discussions with staff and workers in these areas. The inspectors also observed fuel rechanneling operations on the 606' level of the reactor Building. The following observations were made:
A radiation Area sign on the 522' level of the Reactor Building had become undone and fallen to the floor. The rope to which the sign was secured had been taped to the wall and the tape had failed. The
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accompanying Health Physics Foreman resecured the sign.
- An inflated, anti-contamination (anti-c) glove, configured to make
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an obscene gesture, was observed in a controlled area in the Turbine
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Building on top of a Condensate Demineralizer (i.e., a posted radiation area having levels in excess of 2.5 ares / hour).
A large amount of debris, including anti-c gloves, absorbents, plastic bags and tools, was observed in a controlled area on the 467' level of the Radwaste Building in the West Valve Gallery.
- A large quantity of radioactively contaminated tools and equipment was observed to be building up in the Radioactive Material Storage areas.
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Housekeeping throughout the Turbine, Radwaste and Reactor Buildings was observed to be generally poor even for an outage situation.
.The licensee's labeling and posting practices were in compliance with 10 CFR Part 20.203.
These observations were discussed with the Health Physics Supervisor and the Health Physics / Chemistry Manager and at the exit interview.
It was noted that the inflated glove, debris in the West Valve Galley and general poor housekeeping were signs of poor personnel ALARA awareness.
It was also suggested that the institution of a program for contaminated tool reuse could reduce costs for replacement and reduce the quantity of
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radioactive waste being generated. The Assistant Plant Manager stated that he was aware of the housekeeping problem and intended to assign a
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management representative to tour the facilities and initiate corrective action.
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No violations or deviations were identified.
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8.
Followup on Information Notices
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Concerns with the licensee's timely evaluation of IE Information Notices (ins) 83-33, 83-52, 85-06, 85-34, and 85-92 are discussed in paragraph 4 l
of Region V Inspection Report 50-397/86-10. The status of these ins were
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. examined during this inspection. The licensee's progress was concluded
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to be satisfactory.
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No violations or deviations were identified.
9.
Followup on Inspect'or Identified Items (Closed) Followup (50-397/86-10-01). The questions in paragraph 3 of.
I Inspec' tion Report 50-397/86-10 associated with the WNP-2 Semi-Annual Effluent Report dated February 20, 1986, were satisfactorily clarified by the licensee's staff during the inspection. This matter is closed (86-10-01).
No violations or deviations were identified.
10.
Unresolved Item Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, violations or
deviations. Two unresolved items are discussed in paragraph 3.
11.
Exit Interview The inspectors met with the individuals denoted in paragraph I at the conclusion of the inspection on May 2, 1986. The scope and findings of the inspection were discussed.
The unresolved items related to the dosimetry anomalies discussed in paragraph 3 were brought to the licensee's attention.
i The health physics controls associated with the radioactive contamination releases that are discussed in paragraphs 5(a) and 5(c) were discussed at great length. The NRC inspectors stated that even with the time restraints imposed by the schedule that neither event should have occurred if the proper health physics controls had been established. The inspectors emphasized that a good contamination control program is
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essential for maintaining external and internal exposures ALARA.,The inspector added that,a good contamination control program also saves time and reduces costs.
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The inspector also discussed the need for improving the ALARA program and plant cleanliness.
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