IR 05000133/1999001

Insp Rept 50-133/99-01 on 990222-25 & 0308.No Violations Identified.Major Areas Inspected:Decommissioning Performance,Status,Organization,Maint & Cost controls,self- Assessments,Audits & Corrective Actions

ML20205M002
Person / Time
Site:	Humboldt Bay
Issue date:	04/08/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20205L999	List: IR 05000133/1999001 ML20205L995
References
50-133-99-01, 50-133-99-1, NUDOCS 9904150047
Download: ML20205M002 (17)

Text

.

b ENCLOSURE U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

i l

Docket No.: 50-133 License No.: DPR-7 Report No.: 50-133/99-01 Licensee: Pacific Gas and Electric Company (PG&E)

Facility: Humboldt Bay Power Plant Unit 3 Location: 1000 King Salmon A, venue Eureka, California 95503 Dates: February 22 - 25,1999, and March 8,1999 Inspectors: Louis C. Carson 11, Health Physicist Approved By: D. Blair Spitzberg, Ph.D., Chief Fuel Cycle and Decommissioning Branch Attachment: Supplemental Information 9904150047 990408 PDR ADOCK 05000133 G PDR

i

.

-2-EXECUTIVE SUMMARY l Humboldt Bay Power Plant, Unit 3 NRC Inspection Report 50-133/99-01 The Humboldt Bay Power Plant Unit 3 has been in a SAFSTOR decommissioning status since 1976. Based on observations made during the site tour, the facility was being maintained in an acceptable condition. The licensee had continued to monitor conditions related to Unit 3 while '

in SAFSTO Decommissionina Performance. Status. Oraanization. Manaaement and Cost Controls

.

Observations during :ours of the Unit 3 facility and radiologically controlled area indicated that the facuity was being maintained in an acceptable SAFSTOR conditio Fire loading, radiological controls, structural integrity of the facility, and housekeeping were found to be adequate (Section 1).

.

Programs concerning plant status, staffing, and decommissioning cost were reviewed and found to be consistent with the Technical Specifications and the site's SAFESTOR status (Section 1).

Self-Assessment. Audits. and Corrective Actions

.

The safety evaluations reviewed were found to be thorough and in compliance with the j intent of 10 CFR 50.59, which was to identify potential unreviewed safety question !

However, the licensee's 10 CFR 50.59 screening process did not fully consider an issue that conflicted with future decommissioning matters. The licensee had decided that l continuing gross alpha and gross beta groundwater analyses during the  !

decommissioning process was unnecessary. While the deficiency was minor in this case, recognition of such issues are important aspects of the 10 CFR 50.59 process (Section 2). j

.

The licensee's procedure review and approval process was adequate (Section 2).

.

System testing in support of facility design changes had been conducted in accordance with licensee procedures and industry standards (Section 2).  ;

Radiation Protection and Occupational Exposure

. The radiation protection program was adequate to ensure safety during SAFESTOR and i decommissioning activities and met the requirements of 10 CFR Part 20 and Technical Specifications. The health physics program had established effective controls for radioactive material and personnel exposures (Section 3).

,=

.

.

',

-3-Socnt Fuel Pool (SFP) Safety and Cold Weather Preparations

• The SFP water level, and chemistry were found to be in compliance with Technical Specifications. Water clarity and housekeeping in the area around the pool was goo ;

The licensee had maintained a current inventory of the non-fuel llems stored in the SFP. l The freeze protection program was adequate (Section 4).

]

Radioactive Waste Treatment. Effluents. and Environmental Monitorina

. Offsite radiological doses due to gaseous and liquid effluents released from the Humboldt Bay facility in 1997 were reviewed and determined to be within regulatory j limits (Section 5).

Transportation of Radioactive Waste

• The radioactive material shipping procedures and records were reviewed and found to be acceptable and in compliance with NRC and Department of Transportation (DOT) ,

regulations (Section 6). J

• Shipments of radioactive material and the subsequent decontamination and disposition of the material by the licensee's processor were found to be in compliance with NRC regulations and the Humbolt Bay Technical Specification (Section 6).

I e

i

.'

.

L

Report Details 1 Decommissioning Power Reactor inspection Program (2561)

Organization, Management, and Cost Controls (36801)

Decommissioning Performance and Status Review (71801) Inspection Scope This inspection was conducted to verify that Humbolt Bay's programs for implementing SAFESTOR and decommissioning activities were adequate and safe for the work activities in progress. The inspector conducted plant tours, attended planning meetings, and held discussions with licensee's staff regarding the status of the licensee's SAFESTOR and independent spent fuel storage installation (ISFSI) efforts. The licensee's organization was reviewed for any management changes since the last J inspectio .2 Observations and Findinas a. Summarv of Plant Status and Cost Controls The Humboldt Bay Power Plant, Unit 3 was placed in a SAFSTOR decommissioning status in 1976. On February 27,1998, PG&E submitted a Post-Shutdown Decommissioning Activities Report (PSDAR) update to the NRC to comply with 10 CFR 50.82(a)(7). The revised PSDAR described planned changes to the status of the Humboldt Bay Unit 3 facility. Planned changes were reviewed with the licensee and t included: (1) removal of the Unit 3 ventilation stack in 1998, (2) groundwater caisson in leakage repair in 1997, (3) transfer of spent fuel from the SFP to an ISFSI, and (4)

eventual dismantlement of the sit The licensee had essentially completed the removal of the 250-foot ventilation stack and the installation of a smaller stack, ventilation, and exhaust filtration system for Unit The caisson project continued to be successful in that in-leakage was less than 15 gallon / day. However, the licensee was continuing to investigate the source and extent of tritium groundwater contamination around the site (see Section 7.1 for further discussion). The licensee was planning and evaluating options for relocating 390 spent fuc' assemblies from the SFP and storing the fuel in dry cask storage in an onsite ISFS The ISFSI would not be complete until the year 2002, at the earliest. The licensee was evaluating the options for the dismantlement of the SFP once the fuel was removed from the poo J The inspector noted that while the site was officially considered in SAFESTOR status, the projects discussed above were decommissioning activities. Licensee management explained that future decommissioning related projects would require permission from l the public utilities corporation to access decommissioning fundin A tour of the Humboldt Bay facility was conducted as part of this inspection. The facility l appeared in good structural condition. Fire loading and OSHA compliance appeared j l

l

5-acceptable. Good housekeeping practices and radiological controls were observed and determined to be adequate. Tours of the radwaste storage building did not reveal any problems with storage of material. Facility maintenance and upkeep appeared to be adequat b. Oraanization. Staff and Cost The organization for the site is established in Technical Specification, Section Vil(B).

The inspector reviewed the functional organization for the site to include offsite suppor The management organization had remained the same as the last inspection. At the time of this inspection,68 PG&E employees and eight contract personnel worked at the site. The inspector determined that the licensee's staff had been adequate for the SAFESTOR activities that had been conducted to date. However, licensee management had concerns about the need for increasing onsite construction, quality assurance, and engineering in support of the ISFSI projec .3 Conclusions Plant status, organization, staffing, and decommissioning cost were reviewed and found to be adequate. Observations during tours of the Unit 3 facility and radiologically controlled area indicated that the facility was being maintained in an acceptable SAFSTOR condition. Fire loading, radiological controls, structural integrity of the facility, and housekeeping were found to be adequat _10 CFR 50.59 Safety Evaluation Program (37001), and Safety Reviews, Design Changes, and Modifications at Permanently Shutdown Plants (37801)  !

! Inspection Scope The inspector reviewed the licensee's safety review programs associated with the Plant 1 Staff Review Committee (PSRC),10 CFR 50.59 safety evaluations, design changes,  !

facility modification ! Observations and Findinas a. Plant Staff Review Committee Section Vil(D)(1) of the Technical Specifications establishes criteria for conducting the PSRC. During this inspection, the inspector attended one of the licensee's PSRC meetings. Additionally, the inspector reviewed PSRC meeting minutes from 1998 and through February 1999. According to the Technical Specifications,Section VII(D)(1),

the PSRC was to meet at least on a quarterly basis. According to records, the licensee's PSRC met 100 times in 1998. So far during 1999, the PSRC had met seven times. The licensee's PSRC minutes were found to be thorough and reflected that meetings were conducted with a quorum of members, and attemate members as required by the Technical Specifications.

,

. 1

Y-6-

l CFR 50.59 Safety Evaluations

. The inspector reviewed the licensee's Safety Evaluation Procedure HBAP C-19 ;

" Licensing Bases impact Evaluation." The inspector reviewed the seventeen i 10 CFR 50.59 safety evaluations that had been conducted during 1998. All safety evaluations had been reviewed by the PSRC, However, the licensee did not fully consider future decommissioning aspects in one safety evaluatio Safety Evaluation (SE) No.1998-02 concemed the removal of groundwater alpha and beta analyses from the SAFESTOR Offsite Dose Calculation Manual (ODCM). The inspector determined that the licensee's justification for ending alpha and beta groundwater sample analyses was administrative and not technical. The safety evaluation justified that NRC Generic Letter 89-01, which allowed the transfer of radiological and environmental Technical Specifications to the ODCM, did not require alpha and beta groundwater analysis. While the licensee was found to have complied with the applicable procedures, the inspector recommended that the licensee consider the technical merits of alpha and beta groundwater sample analyses during the site's decommissioning process. Licensee management agreed that they had not considered this issue from a technical and decommissioning perspective. The licensee explained that data had consistently demonstrated that alpha contamination did not exist in groundwater around the site, and that tritium was the principal beta contamination contributor. The inspector observed that no data supporting a technical basis for ceasing alpha and beta analyses was included in the original safety evaluatio Licensee management stated that they would continue to collect alpha and beta groundwater samples in support of decommissioning and would implement improvements in future safety evaluation reviews designed to be more sensitive to environmental and decommissioning related matter The safety evaluations reviewed were found to be thorough and in compliance with the j intent of 10 CFR 50.59, which was to identify unreviewed safety questions. However, the licensee's 10 CFR 50.59 screening process did not fully address an issue that conflicted with future decommissioning matters. While the deficiency discussed above was minor in this case, recognition of such issues are important aspects of the 10 CFR ;

50.59 proces J u Procedure Chanaes i The licensee issued revisions to many procedures in 1998. Procedures were reviewed as part of this inspection. Technical Specification Vil(E)" Procedures" required that the licensee establish and maintain procedures for most SAFESTOR activities. Additionally, Technical Specification Vil(L) required that the licensee prepare radiation protection l procedures consistent with 10 CFR Part 20. The inspector found that the licensee L maintained a database of the different types of procedures which included the

'

_

s

I.

> .

l r

r -7-

procedure's responsible department, the latest review dates, and the quality

- classification of the procedure. The inspector reviewed the following procedures: 1

.

Administrative Procedures i l .

Chemical and Radiation Protection Procedures I

! . Radiological Control Standards Procedures )

'

.

Emergency Planning and implementing Procedures i

. Licensing Procedures

.

Maintenance Procedures l . Equipment and Operating Instructions Procedures l . Surveillance Test Procedures

. Security Planning and implementing Procedures in general, changes to the procedures were found to be acceptable and consistent with the current status of the facility and the Technical Specifications. The inspector found that procedure changes had been reviewed and approved by the PSRC and the plant i

'

i

'

manager as required by Technical Specification Vil(D)(1). Out of the 375 procedures that the licensee should review at least once every 2 years, licensee staff had identified that 60 infrequently used procedures had not been reviewed in a timely manne Licensee staff had written a problem report on November 14,1998, to get the overdue procedures reviewed by March 31,1999. The inspector determined that the licensee procedure review and approval process was adequat Plant Desian Chanaes Pursuant to Design Change Package (DCP) M-00429, the licensee had modified and replaced the Unit 3 exhaust fan, ventilation system, and stack. Details leading up to and ,

associated with the construction of the new stack ventilation were explained in NRC !

Inspection Reports 50-133/9803 and 50-133/9804. During this inspection, the inspector l l reviewed the licensee's pre-operational testing of the new stack ventilation system. The inspector reviewed DCP M-00429 and the 10 CFR 50.59 safety evaluation for this plant design change. The inspector held discussions with plant engineering personnel who were responsible for testing the new plant ventilation stack system. The inspector found that the licensee had conducted an extensive test program on the ventilation stack system and the accompanying High Efficiency Particulate Air (HEPA) filter system. The inspector reviewed'the following test procedures:

• SOP 6-1: "HVAC [ Heating Ventilation & Air Conditioning] System Verification Procedure"

. TP 10/1/98: ' Unit 3 Plant Ventilation System Flow Test"

. TP 9/17/98: "New Ventilation System Pre-Operational Checks and Startup Test" The results of the exhaust ventilation flowrate test met the various acceptance criteri The air flow balancing equipment used during the ventilation flow test were found to be in calibration. The HEPA filter was tested in accordance with ANSI N510," Testing of Nuclear Air-Cleaning Systems." The HEPA filter passed the particulate penetration test

.'

.

,

-8-(dioctyi phthalate test). Additionally, the licensee had appropriately calculated the isokinetic sample probe size for the new stack in accordance with ANSI N13.1, " Guide to Sampling Airbome Radioactiva Materials in Nuclear Fecilities."

. The inspector concluded that the licensee's new ventilation stack, HEPA filter, and sampling systems were adequately tested and fully operational.

L Conclusions The safety evaluations reviewed were found to be thorough and in compliance with the intent of 10 CFR 50.59, which was to identify potential unreviewed safety question However, the licensee's 10 CFR 50.59 screening process did not fully consider an issue that conflicted with future decommissioning matters. The licensee had decided that continuing gross alpha and gross beta groundwater analyses during the decommissioning process was unnecessary. While the deficiency was minor in this case, recognition of such issues are important aspects of the 10 CFR 50.59 proces The inspector determined that the licensee procedure review and approval process was adequate. System testing in support of facility design changes had been conducted in j l accordance with licensee procedures and industry standard l 3 Occupational Radiation Exposure; Maintaining Occupational Exposures ALARA; External Occupational Exposure Control and Personnel Dosimetry; internal Exposure Control and Assessment; Control of Radioactive Materials and  !

Contamination, Saveys, and Monitoring (83724,83725,83726,83828, and 83750)

3.1 ' Inspection Scooe  ;

!

The inspector observed work activities to ensure that the radiation protection program !

was implemented and maintained during the SAFESTOR and decommissioning work in !

I- accordance with 10 CFR Part 20 and applicable Technical Specifications. Personnel l exposures and radioactive material control were inspecte < Observations and Findinos a. - Personnel Exposure and As Low As is Reasonably Achievable (ALARA) Goals The inspector reviewed the records system used to track worker exposures. The l l licensee used a combination of thermoluminescent dosimeters'(TLD), pocket chambers, and the access control system to track workers' exposures. Based on TLD records for 1998, the licensee's collective dose was 0.911 person-rem. The individual TLD with the highest exposure during 1998 measured 0.173 rem. So far in 1999, eight individuals had received external exposures which measured 0.064 person-rem collectively. No measured internal exposures were detected during 1998.

t The inspector reviewed the licensee's radiation protection and engineering procedures used to implement ALARA concepts. ALARA reviews were controlled by Procedure RCP 1B," Performing ALARA Reviews for Controlling Occupational Worker

.

f l:

i

!

-9-Exposure." Few significant changes had occurred in the licensee's ALARA program since the facility commenced SAFESTOR activities. Generally, the licensee does not establish site ALARA goals; they set project ALARA goals. The stack removal project ALARA goal was 1.75 person-rem. Accumulated exposures associated with the stack work were 0.671 person-rem. The licensee's rarsiation protection and engineering staffs used ALARA techniques during the planning and implementation of the stack remova The inspector observed that the licensee continued to expend resources on scrabbling and decontaminating concrete and metal associated with the stack removal projec The inspector found that the licensee was unaware of the ALARA concepts that are in Draft Regulatory Guide (RG) DG-4006," Demonstrating Compliance with the Radiological Criteria for License Termination." The inspector showed the licensee that RG DG-4006, Section 3.1.4 had ALARA calculations that specifically addressed concrete scrabbling and decontaminating building surface decontamination Subsequently, the licensee plans to incorporate RG DG-4006 into future ALARA and decommissioning activities. The inspector determined that the licensee's engineering and radiation protection department ALARA design reviews had been conducted adequatel b. Whole-Body Countina and Records The inspector observed a dosimetry technician perform whole-body counter operations, which included the daily energy calibration and quality control chait review. Radiation Control Procedure (RCP) 7S, "Whole Body Counter Operation," set requirements for performing whole-body counter quality control measures such as calibration and ,

'

response checks, and gain checks. The inspector reviewed the whole-body counter check sheets for the months of January and February 1999. The most recent calibration by the vendor was performed on January 14,1998, and the next vendor calibration was scheduled for May 1999. The inspector concluded that quality control checks and calibrations had been performed in accordance with Procedure RCP-7 The inspector reviewed the licensee's process for notifying workers that annual whole-body count were required for unescorted access into radiologically controlled areas (RCA). The licensee requires that individuals must receive a whole-body count annually, at initial site access, and when leaving employment from the site. Based on a review of licensee records, the inspector determined that the licensee met these procedural requirement c. Control of Radioactive Material and Eauioment Release Records The inspector reviewed Humbolt Bay's Procedure RCP-6A," Release of Materials from Restricted Areas" which allowed the removal of tools and equipment from the RCA. The inspector observed stack removal decontamination activities. Surveys of the stack sections had measured contamination that originated from primary to secondary system l leakage. Contamination in the turbine building stack was primarily cesium, cobalt, and tritiuro. Licensee procedures established criteria for release of radioactive material were based on 1,000 disintegrations per minute beta-gamma per 100 cm2 for loose contamination and 100 counts / minute beta-gamma for fixed contamination above background using a pancake probe. However, the licensee used a "No Detectable i s i-10-Release" limits to assure consistency with NRC guidance. The licensee used RG 1.86, l " Termination of Operating License for Nuclear Reactors," and IE Circular No. 81-07,

" Control of Radioactively Contaminated Material.".

l Workers performing decontamination activities were responsible for surveying personal equipment and material before exiting the RCA. Radiation protection technicians were responsible for contamination free release surveys of large equipment and material being removed from the site. Personnel portal monitors, tool monitors, direct pancake probes, and smear surveys were being used adequately. The inspector reviewed the contamination records of tools and equipment being released from the site. Based on the results of the records review, the inspector determined that no tools or equipment were free released by the licensee above the established criteria for release of radioactive material. Records indicated that material that did not meet the free release criteria was shipped to a secondary processor, decontaminated onsite, or stored onsit .3 Conclusions l The radiation protection program was adequate to ensure safety during SAFESTOR and

'

decommissioning activities and met the requirements of 10 CFR Part 20 and the Technical Specifications. The health physics program had established effective controls for radioactive material and personnel exposure Spent Fuel Pool Safety and Cold Weather Preparations (60801,86700, and 71714) Inspection Econe Requirements for the spent fuel pool were reviewed to determine compliance with Technical Specifications ill - VI. A tour of the SFP area was conducted and records of the SFP leakage and freeze protection program were reviewed. This inspection of the l fuel storage building and the SFP areas also assessed housekeeping, fire hazards, SFP material control, and access contro .2 Observations and Findinas a. SFP Inventory The licensee has maintained the spent fuel in the SFP pending the completion of the l ISFSI project. The inspector toured the fuel storage building. The SFP was operational, the pool was covered, and area around the pool was clean. According to records reviewed, SFP water clarity and conductivity were good. Records indicated that the SFP liner, level instrumentation, and other components were in good working conditio Records indicated that materials located in the SFP, around the pool, and around the fuel storage building were appropriately controlled. The inspector reviewed the licensee's records of spent fuelin the SFP which was performed by Procedure STP 3.6.6," Annual Special Nuclear Materials Physical Inventory and Spent Fuel Pool Cover Seal Verification." Additionally, the licensee had identified and documented all non-fuel material stored in the SFP. Movement of non-fuel within the SFP was implemented by

l b

C

~'~

.

J

s-11-Procedure B-3," Movement of Non-Fuel Materialin the Spent Fuel Pool." According to

,

licensee records, the SFP inventory was last completed on August 4,1998, and the last non-fuel inventory was completed on March 17,1994. There were 390 spent fuel assembliss in the SFP that were housed in neutron absorbing cans. Other items that were identified in the SFP included 30 fuel pins, fuel fragments, sample coupons, control rod blades, incore probe parts, control rod drive parts, a 360 Roentgen / hour antimony source, and more. The inspector determinsd that the licensec knew the types r materials in the SFP. However, licensee management indicated that they needed to reassess the condition of the material before its removal from the SFP during decommissionin SFP Monitorina '

SFP surveillance requirements in Technical Specifications, Sections IV and V were

'

'

implemented by several procedures which included Procedure HBAP C-50, " Unit 3 Trending Program", Appendix 8.2, or the SFP trending program. The licensee monitored at:d trended SFP makeup water, liner gap in-leakage, SFP liner gap and pool water radiochemistry, French drain leakage, and groundwater radioactivity. The inspector noted that the licensee monitored the integrity of the spent fuel and maintained SFP water below the Technical Specification limit of 1.0E-4 microcuries/ milliliter cesium 43 The inspector reviewed Procedure STP 3.3.2, " Spent Fuel Pool Level Monitor Verification." Technical Specifications required the SFP water level to be maintained abcve elevation 10 feet 6 inches which is 15 feet above the top of the spent fuel. Daily log sheets were reviewed for the period between November 1990, and January 199 During this period, water level had been maintained at 11.0 feet. The inspector observed that SFP water level was normally monitored from the control room. in addition to operations personnel monitoring water level, annunciator alarms were provided in the con:rol roors for high and low levels. The low level alarm was set for 10 feet 8 inches. The inspector reviewed the SFP level alarm calibration records that were completed on February 18,1999. Records indicated that all SF? level instrumants and alarm switches were operabl SFP Freeze Protection The inspector reviewed the lic ensee's freeze protection program, which was previously documented in NRC Inspection Report 50-133/9801. In general the potential for the SFP cooling system to freeze was minimal. The SFP system is not exposed to open weather and Humbolt Bay is located in the mild climate of Northern California. However, the inspector noted that during short periods in the winter the temperatures would drop below O' Thc inspector reviewed Emergency Procedure EOP-108," Equipment Protection during Extreme Cold Weather." Procedure EOP-108 supported the operating fnssil fuel power plants Units 1 and 2 and the shutdown Unit 3 nuclear piant. The licensee's plan of action for preventing systems from freezing included using electrical heat tracing and draining oiping containing stagnant water. The facility fire system water supply and the

,

-

l

.-

i-12-

'

Unit 3 demineralized water system were identified as the systems that needed cold weather preventive measures implemented to assure that the Unit 3 SFP was saf .3 Conclusion The SFP water !wel and chemistry were found to be in compliance with Technical Specifications. Water clarity and housekeeping in the area around the pool was goo The licensee had maintained a current inventory of the non-fuelitems stored in the SF The freeze protection program was adequat Radioactive Waste Treatment, Effluent, and Environmental Monitoring (84750) and Review of Periodic and Special Reports (90713)

5.1 Insoection Sevoe Data for the 1998 annualliquid and gaseous effluent release reports were reviewe Liquid radioactive waste (LRW) release permits were reviewed to determine compliance with 10 CFR Part 20, Appendix B.10 CFR Part 50, Appendix 1, the Technical Specifications, Section Vl .2 Observations and Findinas Technical Specification, Section Vil (J) required an annual environmental monitoring report and radioactive effluent release report to be issued by the licensee within 90 days !

after January 1 of each year covering the previous 12 months of radioactive effiuent j release activities. At the time of this inspectico, the licensee was working on both j reports to the NRC. However, the inspector reviewed some of the radioactive waste and ;

environmental data from 1998 that will be included in the reports. During 1998 there j were no unplanned radiological releases from the faciiity or releases in excess of ;

10 CFR Part 20, Apperdx (B)(ll) limit Te:.hnical 3pecification, Section Vil(F) and the ODCM set the licensee's requirements an:1 limits for discharpng liquid effluents from the site's waste holdup tanks. Some of the isotopes released during 1998 included tritium (H-3) (0.56 millicuries [ mci]),

strontium-90 (0.083 mci), and cesium-137 (0.04 mci). Considering the volume of dilution water discharged during LRW releases, this amount of radioactivity released equated to less that one percent of the allowable effluent concentration limit. The inspector reviewed records of LRW batch releases from the waste holdup tanks. The concentration of cesiurn-137 and strontium-90 in one batch, prior to being discharged under dilution flowrates, raeasured 82 percent (8.21E-7 micoCi [uciymilliliter (ml]) and 10 percent (5.2E-8 uCi/ml), respectively of the 10 CFR Part 20, Appendix B limit Other isotopes detected in the releases included cobalt-60 and cesium-134. However, the total dose equivalent from these liquid releases were less than one percent of the 10 CFR 20.1301 annual exposure limit for members of the public of 100 millirem.

.

"

The inspector reviewed plant stack gaseous releases for 1998, which included particulate activity releases. Cumulatively, cobalt-60 and cesium-137 releases t:

L

-

,

.

s-13-measured 4.7E-6 Ci and 5.5E-6 Ci, respectively. Tritium releases were estimated at 0.036 Ci. Overall, the total dose equivalent from these releases were less than one percent of the 10 CFR 20.1301 annual exposure limit for members of the public of 100 millire .3 Conclusion The licensee was preparing the effluent report for 1998 as required the Technical Specification. Records demonstrated that through January 1999, radioactivity in liquid

~

and gaseous effluent releases were below the regulatory limit l 6 Transportation of Radioactive Waste (86750)  !

)

6.1 Insoection Scoce The licensee's radioactive material (RAM) shipping program was reviewed for ,

compliance with NRC an ' Department of Transportation (DOT) regulation !

Additionally, the insp' reviched the f:ensee's practice for allowing RAM to be I transferred to seconr' processors for decontamination, burial, and unrestricted l release cf materia !

6.2 Observations and Findinas The inspector reviewed radioactive material shipment records for 1998. Radwaste and I radioactive material shipments were controlled by the following procedures:

- HBRC S-12 Radioactive Material and Waste Shipments

• RCP-6A Release of Liquids, Sludges, Slurries, and Oil From the

[ Radiologically Controlled Area] RCA +

+ RCP-6B Release of Solid Material from Radiologically Controlled Areas

• hdP-61 Collection, Labeling, Packaging, Storage, and Accountability of

Radioactive Material So far in 1999, the licensee had not shipped sny odioactive material offsite. During 1998, the licensee made 21 radioactive material shipments to secondary processors of l

'

radioactive waste. The shipments contained 52 packages of stack concrete, soil, sludge, and metal and represented 33.6 millicuries of radioactivity within 9,050 cubic feet of material. These shipments of radioactive material were classified as limited quantities, low specific activity, and surface contaminated object. The inspector determined that the licensee's shipment records were complete and comprehensiv The inspector had discussions with the licensee concerning the final disposition of the contaminated property that their secondary processor had decontaminated. The inspector also reviewed the contract that the license had with a secondary orocesr,o . )

,

,

.

-14-The inspector reviewed licensee and contractor responsibilities and noted the following about the contract:

The licensee was responsible for the packaging and shipment of the radioactive materia *

Upon receipt of the transferred radioactive material the processor owns i * The processor is responsible for decontamination and volume reductio *

All material shall be decontaminated to the processor's unrestricted release limit =

Any material t.iat does not mat the contractors release limits shall be disposed of at a licensed radioactive waste disposal facility, returned to the licensee, or shipped to another processo =

In accordance with NRC RG 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Waste and Release of Radioactive Material Liquid and Gaseous Effluent from Light Water Cooled Nuclear Power Plants," the contractor had to provide the licensee a summary report that included radioactive waste volume, weight, and activity of Humbolt Bay Unit 3 material that had been decontaminated and shipped to a low-level waste facility by the processo At the time of this inspection, licensee processors had reported in accordance wiih RG 1.21 that 550 cubic feet of former Humbolt Bay material containing 9.76 millicunes of radioactivity had been buried at a low-level waste facilit l 6.3 Conclusions The radioactive material shipping procedures and records were reviewed and found to be acceptable and in compliance with NRC and DOT regulations. Shipments of radioactive material and the subsequent decontamination and disposition of the material l

'

by the licensee's processor were found to be in compliance with NRC regulations and the Humbolt Bay Technical Specification Follow-up of Open items (92701)

l 7.1 (Discussed) Inspection Followuo item (IFI) 50-133/98001-03 Tritium in Monitorina Wells The licensee was continuing to assess the potential source for the tritium detected in groundwater monitoring wells. Samples inken February 1998, found tritium levels of 274 picocuries/ liter (pCi/l) and 447 pCi/l. This compared to samples taken November 18-20,1997, which indicated 312 pCi/l for Well No.11 and 401 pCi/l for Well No.1. In 1987 tritium in monite gnwells had measured E,pproximately 2,700 pct /!.

Humboldt Bay Technical Sp a stion, Table V-1 established a reporting criteria for i

. _

I

l

,'

,

l l

l -15-i tritium of 30,000 pCi/l. For drinking water, the Environmental Protection Agency limit for tritium is 20,000 DCi/l.

i  !

l The tritium c . continued to decrease over the years at a rate faster than natural decay (the half life of tritium is 12.3 ycais). However, the tritium should have dissipated more l quickly due to the transportability of tritium in ground water. This raisea the concern that a source of tritium may be present near the monitoring wells. Tritium found in the r:atural environment is approximately 10 pCi/l and would not account for the levels found in the moaitoring wells. The largest tritium source onsite was the spent it;ai poo Tritium levels in the pool were measured at 98,000 pCi/l in February 199'/.

The inspector reviewed a licensee memorandum dated February 19,1999 fhe licensee had proposed to investigate the source and extent of tritium contamination in groundwater in the vicinity of Humbolt Bay. The results of this investigation will be i reviewed during a future inspection.

i i

,

8 Exit Meeting

,

The inspector presented the inspection results to members of the licensee management l

at the exit meeting on February 25,1999, and during a subsequent telephone l conversation on March 8,1999. The licensee acknowledged the findings presente The licensee did not identify as proprietary any information provided to, or reviewed by, the inspector.

l l l l \

i l

l l

!

I l

I

l

!

<

-

l a

.

s AT, "HMENT PARTIAL LIST OF PEi.MNS CONTACTED Licensee J. Albers, Senior Radiation Protection Engineer (Radiation Protection Manager)

L. Claytor, Radwaste Engineer J. Crow, Training M. Grossman, Operations Supervisor E. Kahler, Decommissioning Project Manager T. Moulia, Plant Manager R. Parker, Radiation Protection D. Sokolsky, Senior Licensing Engineer j R. Willis, Plant Engineer INSPECTION PROCEDURES USED IP 2561 Decommissioning Power Reactor inspection Program IP 36801: Organization, Management, and Cost Controls IP 37801: Safety Reviews, Design Changes, and Modifications l

IP 40500: Effectiveness of Licensee Controls in Identifying, Resolving and Preventing

-

Problems IP 40801: Self-Assessments, Audits, and Corrective Actions l IP 60801: Spent Fuel Pool Safety IP 71714: Cold Weather Preparation IP 71801: Decommissioning Performance and Status Rev:ew IP 83750: Occupational Radiation Exposure IP 83724 External Occupational Exposure Control and Personnel Dosimetry IP 83725 Internal Exposure Control and Assessment IP 83726 Control of Radioactive Materials, and Contamination, Surveys, and Monitoring IP 84750 Radioactive Waste Treatment, Effluents, and Environmental Monitoring IP 86700 Spent Fuel Pool Activities IP 86750 Radwaste and Transportation IP 92701 Followup on Open items l

i- .

$

2-i

)

LIST OF ACRONYMS j ALARA as low as is reasonably achievable ANSI American National Standards institute CFR Code of Federal Regulations

• C degrees Celsius DCP design change package DSAR Defueied Safety Analysis Report ISFSI independent spent fuel storage installation IFl inspection followup item LRW liquid radwaste mci millicurie NRC Nuclear Regulatory Commission PG&E Pacific Gas & Electric pCi/l ptocuries/ liter PSDAR Post-Shutdown Decomm:ssioning Activities Report PSRC plant safety review committee 1 RP radiation protection  !

RAM radioactive material TLD Thermoluminescent Dosimeter uCi microcurie ITEMS OPENED, CLOSED, AND DISCUSSED Ooened i None Closed None piscussed 50-133/9801-03 IFl Tritium in Monitoring Wells i