ML20214L980
| ML20214L980 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 04/01/1986 |
| From: | Wenslawski F NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Scarano R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| Shared Package | |
| ML20214L963 | List: |
| References | |
| TAC-64735, TAC-64736, NUDOCS 8609100488 | |
| Download: ML20214L980 (2) | |
Text
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UNITED STATES
- 8 1 NUCLEAR REGUE.ATORY COMMISSION e'
{ ,I REGION V g 1450 MARIA LANE. SUITE 210 0,%
- ,d WALNUT CREEK, CAllFoRNIA 94593 April 1, 1986 liEMORANDUM FOR: Ross A. Scarano, Director Division of Radiation Safety and Safeguards FROM: Frank A. Wenslawski, Chief Emergency Preparedness and Radiological Protection Branch
SUBJECT:
MINUTES, NRR MEETING, OFFSITE CONTAMINATION RESULTING FROM RANCHO SECO LIQUID EFFLUENTS, 03/25/86 Following is my perception of the proceedings of the subject meeting:
Attendees:
NRR IE ELD RV EPA John Stolz Ed Flack Steve Burns George Kalman Frank Wenslawski Allen Richardson John Buchanan Linda Gilbert Mike Wrangler Roger Pederson Charles Willis Frank Congel Ed Brannagan Jerry Swift Charles Nickles The meeting was opened by John Stolz. He was followed by George Kalman who provided an overview of the history of the offsite contamination problem at Rancho Seco. The meeting was then opened to discussion of the points identified in Kalman's memorandum dated March 18, 1986 (enclosed).
Considerable discussion pursued with the intent to have a common understanding of the requirements of 40 CFR 190, Rancho Seco Technical Specifications and the best course of action to take with the licensee. The following was agreed upon:
The EPA representative stated and NRC representatives agreed that it is the intent of 40 CFR 190 to consider the total dose from all sources, including that resulting from an environmental buildup of radioactivity, when determining compliance with the 25 mrem annual dose limit.
Charles Willis stated, without significant opposing opinions by other attendees, that the standard Technical Specifications issued to implement Appendix I to 10 CFR 50 (3 mrem / year total body liquid
- pathway dose) were not intended to consider dose from an environmental buildup of radioactivity. The 3 mrem limit applies only to releases occurring within a year. It was generally agreed that the wording of the Technical Specifications is not specific regarding this intent.
8609100488 860829 PDR ADOCK 05000312 p PDR
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. Ross A. Scarano April 1, 1986 Actions agreed upon are:
(1) NRR will ask the licensee to submit a technical specification change to decrease the LLD to provide sufficient sensitivity to determine compliance with the Appendix I implementing technical specification.
(2) NRR will ask the licensee to submit a plan to ensure no future exposure in excess of 40 CFR 190 limits will occur.
(3) Technical Specification factors to receive further review by NRR include: an acceptable LLD; a clarification that the 3 mrem /10 mrem limits of Technical Specification 3.17.2 apply only to releases occurring during a year; a clarification that Technical Specification 3.25 will require an annual determination of compliance with 40 CFR 190, including dose from existing environmental radioactivity and possibly a projection of future doses; possible Technical Specification changes to more precisely define land use census and environmental monitoring program to assure continued compliance to 40 CFR 190.
(4) The issu'e of environmental contamination at Rancho Seco is one of low level dose and is not considered a safety issue. Therefore, any Technical Specification actions do~not justify an order and such actions will be pursued through the normal licensing channels.
NRR estimates six months to accomplish Technical Specification changes.
(5) NRR agrerd to determine the licensee's plans for construction of evaporation ponds.
(6) NRR will eventually write a letter to EPA identifying the actions NRC has taken on the Rancho Seco offsite dose issue.
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(7) Region V agreed to conduct an inspection to determine the degree, if any, of licensee's manipulation of the liquid LLD for releases and the consistency of the LLD used for releases with that reported in the semiannual report. The possibility of LLD manipulation was originally discussed between a SMUD representative and NRR, and subsequently between a SMUD representative and Region V.
rank A. Wenslawski, Chief Emergency Preparedness and Radiological Protection Branch
Enclosure:
as stated cc w/ enc: J. Stolz, PBD6, NRR G. Kalman, PBD6, NRR J. Martin, RV R. Scarano, RV G. Yuhas, RV l L. Miller, RV l
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July 22,1896 5 o
,. Docket No. 50-312 Mr. John E. Ward -
Assistant General Manager, Nuclear Sacramento Municipal Utility District 6201 S Street P. O. Box 15830 Sacramento, California 95813
Dear Mr. Ward:
As stated in my letter of April 28, 1986, the NRC staff has been reviewing the Rancho Seco Radiological Effluent Technical Specifications (RETS). To date, this review has been limited to pertinent aspects of the requirements for effluent monitoring, nearby land use identification, offsite dose
, calculations, environmental monitoring, and limitations on calculated offsite deses. This review includes the Rancho Seco Offsite Dose Calculation Manual
~ (ODCM), and addresses the degree to which compliance with this document, along with the RETS, provides assurance of meeting the requirements of 10 CFR Part 50, Appendix I, and 40 CFR Part 190 concerning radiation doses received by members of the public.
The above review was undertaken because of the conclusions of the February 24, 1986 document, "NRC Assessment of the Environmental Radioactive Contamination in the Vicinity of the Rancho Seco Nuclear Power Plant," a copy of which was transmitted to you by my April 28, 1986 letter. The above assessment noted that radionuclide concentrations near the Rancho Seco site resulting from the release of liquid effluents from the plant were found to be at levels which could cause maximum-exposed individuals to receive potential doses well in excess of the RETS dose limits based on the regulations cited. This brings to question whether the Rancho Seco RETS and ODCM provide adequate assurance that the requirements of the regulations are being met.
Specifically, as a result of the NRC Staff's review undertaken in connection with contamination found in the vicinity of the Rancho Seco Plant, the NRC staff found that there is an inconsistency between the Lower Limit of Detection (LLD) as listed in Table 4.21-1 of the Technical Specification Surveillance Standards (Section 4.21.1) and the Technical Specification sections (3.17.2 and 4.21.2) relating to 10 CFR Part 50, Appendix I design objectives. Because of the highly atypical characteristics of the cooling water system and of the receiving waters, liquid effluent releases at the present LLD values, if properly calculated, can result in doses in excess of Part 50 Appendix I and the limits specified in 40 CFR 190.
To correct this situation and to assure that offsite releases are in accordance with applicable Federal Regulations, you should propose appropriate changes to the Rancho Seco Technical Specifications.
4 c p a s m>> n
4 Mr. John E. Ward o After consideration of our enclosed evaluation, we suggest that a mceting or conference call be held with the staff to discuss your future plans for dealing with the concerns indicated and for assuring continued compliance with the regulations.
Sin,cerely,
-= L-.. 4,; m y J% p, ym, ,
John F. Stolz, Director PWR Project Directorate (6 Division of PWR Licensing-B
Enclosure:
As Stated I
cc w/ enclosure:
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- See previous white for concurrences.
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STAFF EVALUATION OF RANCHO SEC0 RADIOLOGICAL EFFLUENT '
TECHNICAL SPECIFICATIONS AND OFFSITE DOSE CALCULATION PANUAL REGARDING OFFSITE DOSES FROM LIQUID EFFLUENT Concerns have recently been raised regarding possible unacceptable radiation levels in liquid effluent releases from the Rancho Seco plant. The following is a staff evaluation of these concerns based in part on the conclusions con-tained in the February 24, 1986 document, "NRC Assessment of the Environmental Radioactive Contamination in the Vicinity of the Rancho Seco Nuclear Power Plant." Our evaluation considered the current Rancho Seco Radiol'ogical EffluentTechnicalSpecifications(RETS)andOffsiteDoseCalculationManual(OD'CM) which are intended.to assure that doses from liquid effluent releases are within the limits specified'in the regulations.
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NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," provides calculational models for dose contributions for implementing 10 CFR Part 50, Appendix I "as low as is reasonably achievable" requirements. For liquid effluent releases, a near field average dilution factor is used which takes into account the maximum undiluted liquid waste flow,
{ the combined liquid releases for each unit, and the mixing effects in the receiving water body in the near field of the discharge structure. For plants with non-recirculating main condenser cooling systems, the mixing effects in the receiving water body are ignored for conservatism. However, for plants with recirculating cooling systems, where cooling water discharge flowrates are much less than for plant with non-recirculating cooling systems, credit is allowed for mixing effects in the near field of the receiving water body up to the degree of dilution that would be achieved by a 1000 cubic feet per second flow rate of non-recirculating cooling water.
Rancho Seco has a recirculating main condenser cooling system. Based on a comparison of Environmental Statements for various nuclear power plants, the Rancho Seco design average discharge flow rate is one of the lowest of all l U. S. nuclear power plants. As with similar plants, the liquid waste discharge
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includes condenser cooling and service water system blowdown, and ot[her minor streams in addition to liquid radwaste effluents. However, atypically, at Rancho Seco there is little or no dilution of liquid wastes after discharge from the plant discharge structure due to the almost total absence of a receiving water body comprised of water other than from the plant discharge.
Consequently, no credit is provided in the Rancho Seco ODCM for mixing in the receiving water body in the near field of the discharge structure.
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The Rancho Seco RETS are patterned after the guidance of NUREG-0472, " Standard Radiological Effluent Specifications for Pressurized Water Reactors," which is in turn based on typical reactor plant designs and site characteristics. Both
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the Rancho Seco RETS and NUREG-0472 incorporate provisions for:
- 1. Radioactive liquid effluent monitoring instrumentation and sampling to
, monitor and control the release of radioactive material during actual or potential releases;
- 2. ' An annual land use census to ensure that changes in the use of areas at and beyond the site boundary are identified and modifications to the environ-mental monitoring program are made if required by the results of the census;
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- 3. An ODCM to contain the methodology and parameters to be used in the cal-culation of offsite doses due to radioactivity in effluents;
- 4. An environmental monitoring program which supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of environmental exposure pathways; and
- 5. Limitations on calculated doses to members of the public to implement 10 CFR Part 50, Appendix I, and 40 CFR Part 190.
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The Rancho Seco RETS provide for certain radioactive liquid effluent monitoring instrumentation to monitor and control, as applicable, releases of radioactive material during actual or potential releases. Each batch of liquid waste from the waste release tank must be sampled and analyzed prior to release for the principal gamma emitters (Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134, <
Cs-137, Ce-141 and Ce-144) and I-131; one batch sample per month is analyzed for dissolved and entrained gases (gamma emitters); and monthly or quarterly composites of the batch samples are analyzed for tritium, gross alpha, Sr-89
, and Sr-90. A gross radioactivity monitor is required to be operable on the regenerant hold-up tank discharge line.
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To comply with the Rancho Seco RETS, the lower limit of detection (LLD) for these radionuclides must be'as follows: 5 x 10~7 microcurie per milliliter or better for the principal gamma emitters, 1 x 10~0 microcurie per milliliter or
, better for I-131, 1 x 10-5 microcurie per milliliter or better for dissolved and entrained gases and for tritium, 1 x 10~7 microcurie per milliliter or better for gross alpha, and 5 x 10-8 microcurie per milliliter or better for Sr-89 and Sr-90, which is in accordance with with the guidance of NUREG-0472.
NUREG-0472 LLD's, however, are based on typical plant designs and site -
characteristics. By definition in the Rancho Seco ODCM, the LLD is the smallest concentration of radioactivity in a sample which will be detected and reported as a positive value approximately 95 percent of the time. Conversely, a sample with no real net activity above background will be reported as a positive value about 5 percent of the time. The LLD is a predictive estimate representing the capability of the measuring system, not an after-the-fact estimate for a particular sample.
1 As reported in the Rancho Seco, " Radiation Exposure Environmental Protection, Effluent and Waste Disposal 1984 Annual Report," there were no liquid effluent releases for that year with detectable levels of Strontium-90, Cobalt-57, Zinc-65, Chromium-51, Iron-59, Molybdenum-99, Technetium-99m, Barium-T Lanthanum-140, and Cerium-141. Regulatory Guide (RG) 1.21. " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radio-ative Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear
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Power Plants," provides that, if radioactivity in the sample is less than the l
detection capability of the instrument, then the value shculd be reported as less than the sensitivity of measurement (the numerical value is to be t indicated). It is not clear from the Rancho Seco documentation reviewed whether the licensee used, for the purpose of calculating the reported offsite doses, the actual LLD values, or lesser values for radionuclides in effluent measurements where concentrations in excess of the LLD values were not detected.
The staff has calculated, based on the Rancho Seco ODCM,* the annual offsite I
doses which would be attributable to each radionuclide if released at Rancho .
Seco in waste water at the LLD activity concentration specified in the Rancho Seco RETS. Based'on the information provided in the Rancho Seco, " Radiation Exposure, Environmental Protection, Effluent and Waste Disposal 1984 Annual Report," it was assumed that the waste water is diluted 100-fold prior to discharge offsite. The calculated annual adult total body dose is about 79 millirems, with Cesium-134, Cesium-137 and Strontium-90 contributing about i
40, 25 and 7 millirems, respectively. The calculated annual adult bone dose is about 82 millirems, with Strontium-90, Cesium-137 and Cesium-134 contributing i
about 26, 28 and 21 millirems, respectively. The calculated annual adult liver
, dose is about 101 millirems, with Cesium-134 and Cesium-137 contributing about 49 and 37 millirems, respectively. The calculated annual adult thyroid dose is a'oout 41 millirems, with Iodine-131 contributing about 37 millirems.
In contrast, Rancho Seco Technical Specification 3.17.2, which is provided to implement the requirements of 10 CFR Part 50, Appendix I, requires that the annual dose to a member of the pub 1'ic from radioactive materials in liquid
- Revision 3, dated 09-23-85, provided as Attachment 8 to Rancho Seco, " Effluent and Waste Disposal Semiannual Environmental Report, January to June 1985."
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effluents be limited to 3 millirems to the total body and to 10 millirems to any organ. Rancho Seco Technical Specification 3.25, which is provided to meet the dose limitations of 40 CFR 190, requires that the annual dose to a member of the public due to releases of radioactivity and radiation from fuel cycle sources be limited to 25 millf rems to the total body or any organ (except the thyroid, which is limited to 75 millirems).
Clearly, the incorporation of the model RETS LLD values in the Rancho Seco RETS
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is an error since use of these values can result in releases of radioactive materials.to which offsite doses may be attributed (through the use of the methodology of the ODCM) that are in excess of the limits provided by the RETS to implement the regulations,10 CFR Part 50, Appendix I, and 40 CFR Part 190.
To the extent that'a significant dose' contribution may be present at Rancho Seco due to the release of radioactive materials in concentrations below the LLD values, and dose calculations based on these releases do not adequately account for possible radioactivity in these releases, the potential may exist for concentrations of radioactive materials and levels of radiation in the environment that are higher than expected on the basis of the effluent measurements, and possibly exceeding regulatory limits.
The Rancho Seco RETS, utilizing the guidance of NUREG-0472, provide for an annual land use census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitor-l ing program are made if required by the results of the census. To the extent that significant changes in the use of nearby areas may occur and remain for significant periods of time until identified 'in the annual land use census, a potential may exist for concentrations of radioactive materials and levels of radiation in the environment to be higher than expected on the basis of the previous land use census, and to possibly exceed regulatory levels. Such changes may have a greater impact for Rancho Seco when compared to the more typical model plants because of the differences in site characteristics.
Thus, a more frequent land use census may be necessary.
l The Rancho Seco 0DCM is based on and closely follows the guidance provided in RG 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, ,
Appendix I." RG 1.109 is also based on typical reactor plant designs and site characteristics. The basic features of the calculational models described in '
RG 1.109 are general approaches that the NRC staff has developed for appli-cation in lieu of specific parameters for specific sites. RG 1.109 encourages the use of site-specific values, assumptions and methods.
Calculational methods for doses due to liquid effluents are provided in RG
- 1.109 for the pathways of potable water, aquatic foods, direct radiation from .
shoreline deposits, and foods grown on land irrigated with contaminated water.
The calculational model provided in RG 1.109 for the aquatic foods pathway assumes that the concentrations of radionuclides in aquatic foods are directly related to the concentrations of the radionuclides'in the water. The model, therefore, does not take into account the effects of the long-tem buildup of concentrations of the radionuclides in bottom sediments on the doses due to ingesting aquatic foods (bottom-feeding fish). However, for purposes of cal-culating annual direct doses from shoreline deposits, the models of RG 1.109, do take into account the long-tem buildup of radionuclides in sediment due to the transfer of the radionuclides from the water to the sediment over the facility operating life. Further, for the purpose of calculating annual doses from consumption of foods grown on irrigated land, the long-tem buildup of radionuclides in soil is also accounted for. Calculational models are not provided for doses from liquid effluent releases due to swiming and due to direct radiation from the long-term buildup of radionuclides on land irrigated with contaminated water.
The Rancho Seco ODCM, Revision 3, states that the models account for all potential land, water usage, and food radiological exposure pathways that could actually exist downstream from Rancho Seco. However, the models do not take into account
! the effects of the long-tem buildup of concentrations of radionuclides in r
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O bottom sediment on doses due to ingesting aquatic foods, or direct radiation from the long-tem buildup of radionuclides on land irrigated with contaminated water. Therefore, to the extent that a significant dose contribution may be present at Rancho Seco through the omitted pathways, a potential may exist for concentrations of radioactive materials and levels of radiation in the environment that are higher than expected on the basis of the existing modeling of the environmental exposure pathways, and for these levels to possibly exceed regulatory limits.
The Rancho Seco RETS Environmental Radiological Monitoring Program provides for the collection and analysis of specifed numbers of samples of surface water, runoff water, shoreline mud and silt, milk, fish, and several classes of r
harvested food at specified' frequencies. This program is based on the guidance of the model program in NUREG-0472. The sampling locations are described in the l
ODCM. The RETS provide reporting levels for radionuclide concentrations in the environmental samples in order to appropriately identify when concentrations of radioactive materials and levels of radiation may be higher than expected on the
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basis of'the effluent measurements and the modeling of the environmental exposure pathways.
Based on the above infomation, we conclude that the situation at Rancho Seco involving the availability of only minimum quantities of cooling and receiving water for dilution of liquid effluents is highly atypical. Therefore there is a greater likelihood that offsite doses will approach the dose limits when compared to the more typical plants envisaged by the regulatory standards with similar waste characteristics and waste treatment to that at Rancho Seco. The Rancho Seco RETS provide requirements for effluent monitoring, nearby land use -
identification, offsite dose calculations, environmental monitoring, and limitations on offsile doses. The RETS and the ODCM are patterned after_
model requirements which are based on more typical plant designs and site characteristics. Therefore, to provide assurance that plant operation maintains releases below the dose limits, atypically greater levels of accuracy or conservatism may be needed in quantifying liquid releases, identifying nearby
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. land uses, and calculating offsite doses at Rancho Seco. Further, an error exists in the Rancho Seco RETS LLD values which can result in doses attribut-able to liquid effluent releases that are in excess of the RETS limits for maintaining compliance with the requirements of 10 CFR Part 50, Appendix I and 40 CFR Part 190.
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UNITED STATiiS
[ g, NUCLEAR REGULATORY COMMISSION s 8 REGION V f 1450 MARIA LANE.SulTE 210
%, ,e WALNUT CREE K, cALIFORNI A 94596 JUN 061986 .
Docket No. 50-312 Sacramento Municipal Utility District P. O. Box 15830 Sacramento, California 95813 Attention: Mr. John E. Ward Assistant General Manager, Nuclear Gentlemen:
Subject:
NRC Inspection Report This refers to the special inspection conducted by Messrs. G. Basada and ,
G. Tuhas of this office on April 1, 2, 29, and May 15, 1986, of activities authorized by NRC License No. DPR-54, and to the discussions of our findings held by Mr. Yuhas with Messrs. R. Colombo and R. Rodriguez and other members
, of your staff at the conclusion of each site visit during the inspection.
This inspection was conducted.to evaluate your management of liquid radioactive effluents during 1985. The inspection consisted of selective
====fnations of procedures and representative records, interviews with personnel, and observations by the inspectors.
Enforcement action related to the enclosed inspection report will be addressed in separate correspondence.
In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure will be placed in the-NRC -Public Document Room. -
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Sacramento Municipal Utility District Should you have any questions concerning this inspection, we will be glad to discuss them with you.
Sincerely, Udgbr si n:1 ey F. A Wcnsinici Ross A. Scarano, Director Division of Radiation Safety and Safeguards
Enclosure:
Inspection Report No. 50-312/86-15 cc w/ enclosure: .
L. G. Schwieger, SMUD G. Coward, SMUD State of CA FEMA, Region II bec: RSB/ Document Control Desk (RIDS)
Resident Inspector Project Inspector G. Cook /B. Faulkenberry/J. Martin bec w/o enclosure:
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RScarano 0 $f '
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U. S. NUCLEAR REGULATORY COMMISSION REGION V Report No. 50-312/86-15 Docket No. 50-312 License No. DPR-54 Licensee: Sacramento Municipal Utility District P. O. Box 15830 Sacramento, California 95813 Facility Name: Rancho Seco Nuclear Generating Station Inspection at: Clay Station and Sacramento, California Inspection conducted: April 1, 2, 29, May 15 and subsequent telephone discussions through May 23, 1986 Inspectors: h (A M _ (m/f/$$
G. Hama Radiation Laboratory Specialist Dat'e Signed 0@ Uk G. P. Y a , Chief, Facilities Radiological 6/A N6 Date Signed etion S ction Approved by: '.[ f g/Mf/
F.'A. Dat'e signed
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Wenslawski, Chief, Emergency Preparedness and Radiological Protection Branch Summary:
Inspection on April 1, 2, 29, and May 15, 1986 and subsequent telephone discussions through May 23, 1986 (Report No. 50-312/86-15)
Areas Inspected: Special unannounced inspection by two regionally based NRC specialists to close previously identified Unresolved Item 50-312/84-06-01, and to review the licensee's management of radioactive materials released in liquid affluents during 1985. The following Inspection Procedures were utilized: 30703, 84523, 84723, 84725, 92700, 92701, and 90713.
Results: Of the threa' areas inspected, apparent violations involving failure to develop procedures to implement 10 CFR 50, Appendix I criteria and f ailure to report the results of radioactivity measured in liquid effluent (Paragraph 4); failure'to comply with T.S. 3.17.2 liquid effluent dose limits for 1985 (Paragraph 5); failure te perform safety evaluations required by 10 CFR 50.50 and failure to establish, implement and maintain procedures required by T.S. 6.8 (Paragraph 6) were identified.
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Details
- 1. Persons Contacted A. Licensee Personnel
+R. Rodriguez, Assistant General Manager, Nuclear
- *R. Powers, Manager, Nuclear Engineering
- J. McColligan, Assistant Manager, Nuclear Plant
- S. Redeker, Manager, Nuclear Operations
- F. Kellie, Radiation Protection Superintendent
- R. Colombo, Regulatory Compliance Supervisor
+*E. Bradley, Supervising Health Physicist
- C. Stephenson, Principle Regulatory Compliance Engineer D. Mixa, Cost Analfst
- B. Wilson, Senior Chemist'ry and Radiation Assistant (SCRA) ,
- S. Manofsky, SCRA W. Hampton, Chemistry and Radiation Protection Technician (CRPT)
D. Kearl, CRPT M. Leiwander, CRPT W. Partridge,'CRPT ,
B. Non-Licensee Personnel R. Miller, Acting Chemistry Supervisor, Sierra Technology R. Gardner, Certified Health Physicist, United Energy Services Corp.
R. Oesterling, Certified Health Physicist, United Energy Services Corp.
t C. Nuclear Regulatory Commission (NRC)
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- +G. Perez, Acting Senior Resident Inspector
- Denotes attendance at exit interview conducted on April 2, 1986. ,
+ Denotes attendance at exit interview conducted on April 29, 1986.
- Denotes attendance at exit interview conducted on May 15, 1986.
In addition to the individuals identified above, the inspectors met with contractors and other members of the licensee's staff.
2.- Unresotved Item (50-312/84-06-01)
NRC Inspection Report 50-312/84-06, dated May 31, 1984, describes an NRC Region V concern that members of the public may have received a dose from ,
ionizing radiation in excess of the values presented in 10 CFR 50 Appendix I, Technical Specification objectives and 40 CFR 190 as a result of radioactive materials contained in liquid effluents released from the Rancho Seco Nuclear Generating Station (RSNGS).
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In a special report dated May 14, 1984, the licensee provided the results of calculations which indicated chat these values had been exceeded for l 1981, 1983, and 1984 for a hypothetical " maximum adult" exposed via the liquid-fish-man pathway. The licensee stated that based on the concentrations measured in fish flesh and a whole body count of the l " maximum" individual, the actual calculated dose to a real member of the l public was 12 arem and therefore they had not exceeded the 25 arem per year standard of 40 CFR 190.
Region V requested the NRC Office of Nuclear Reactor Regulation (NRR) to establish the validity of the licensee's calculations and to determine if the values presented in 40 CFR 190 had been exceeded.
NRR's evaluation included review of numerous licensee reports, an extensive environmental survey performed by Oak Ridge National Laboratory (NUREG/CR-4298) and aerial measurements of radioactive materials in the vicinity of the RSNGS performed by EG&G Energy Measurements, Inc. NRR's evaluation was completed in the spring of 1986 and the results l transmitted by separate correspondence to the licensee and Region V.
NRR found that during 1984 (the most limiting year) it could not be ,
concluded that the whole body dose to the maximally exposed member of the public as determined from environmental measurements exceeded the 25 area
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standard of 40 CFR 190 in view of whole body count data. The calculated dose to this real person based on measured radionuclide concentrations in fish flesh and recalled ingestion rates was about 50 area. However, a whole body count performed on the individual failed to detect any radioactivity associated with releases from RSNGS. The whole body count I had a minimum detectable activity which would have confirmed a dose of about 7 area.
Accordingly, since it has not been reasonably established that a real member of the public received a dose in excess of the 40 CFR 190 i standard, no violation of 10 CFR 20. 106(g) has been identified. This matter is closed.
l 3. Radioactive Liquid Effluents During 1985 A. Background On July 21, 1984, the licensee implemented Amendment No. 53 to the RSNGS Technical Specifications (T.S.). This amendment incorporated 10 CFR 50 Kppandix I, Numerical Guides for Design Objectives and i Limiting Conditionc for Operation to, Meet the Criterion "As Low As
- Is Reasonably Achievable" for Radioactive Material in Light-Water Cooled Nuclear Power Reactor Effluents, requirements into the T.S.
The licensee submitted Special Report No. 84-07, on September 27, 1984 (RJR 84-425) as required by T.S. 3.17.2 and 3.25 to report that
- the cumulative calculated radiological exposure resulting from ,
I liquid effluents exceeded the calendar quarter and~ calendar year dose limits of T.S. 3.17.2 and fuel cycle dose limit of T.S. 3.25 for calendar year 1984 through August 31, 1984. In their September 27, 1984 letter, they stated that, "The District now is limiting its discharges so that 10 CFR 50 Appendix I limits will not
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- be exceeded." In their " Attachment to Special Report 84-07" the licensee presented their near and long term corrective actions. In response to this licensee sube*ttal, NRC issued a letter dated November 15, 1984 which concluded, based on the corrective actions taken and planned, that a variance for continued operation pursuant to 40 CFR 190.11 was not needed.
During follow-up inspections conducted in November 1984 and October 1985, Region V inspectors found the licensee was implementing the near term corrective action involving the Polishing Demineralizer System and the Regenerant Holdup Tanks (RHUT) (Inspection Report No. 50-312/84-27) and that review of the liquid effluent release records confirmed no detectable concentrations of fission a*
activation products were apparently released as descrited in the licensee's Semiannual Effluent Radioactive Release Report dated September 26, 1985 (Inspection Report No. 50-312/85-28).
The licensee's contractor, Lawrence Livermore National Laboratory (LLNL) collected two water samples from the Rancho Seco RHUTs on October 14-15, 1985, for isotopic analysis. On November 22, 1985, the licensee's Supervising Health Physicist discussed the adequacy of Rancho Seco's lower ^ limit of detection (LLD) capability in terms of the 10 CFR 50 Appendix I criteria with the NRC Region V, Chief, Facilities Radiological Protection Section. The licensee representative stated that he had initiated a study of the matter and samples had been taken. The Supervising Health Physicist advised the Chief. Facilities Radiological Protection Section, on November 26, 1986, that based on verbal results from LLNL; radioisotopes of cesium and cobalt had been detected at l concentrations about a factor of two below the onsite Rancho Seco laboratory capability. Since extrapolation of this one data set for all liquid releases made during 1985 could call into question
- compliance with the 3 arem per year total body and 10 arem per year l organ dose commitment of Appendix I and T.S. 3.17.2, other possible l sources of data were discussed. The licensee representatives
[
. indicated that composite samples of each liquid batch release collected pursuant to T.S. 4.21.1 and sent to another contractor, Controls For Environmental Pollution, Inc. (CEP), could be analyzed for gamma emitting isotopes with an LLD better than the onsite capability. The licensee pointed out that for cesium and cobalt T.S. Table 4.21.1 requires an LLD of SE-7 uCi/ml, the onsite laboratory reported typical LLDs of SE-8 uCi/ml, CEP ranged from 2E-9 to 1EJ8 uC1/ml while LLNL reported values of 2E-11 uCi/ml for their LLD.
l f The licensee defines LLD'in their Offsite Dose Calculation Manual I
. (OBCM) as:
i "The smallest concentration of radioactivity in a sample which will be detected and reported as a positive value approximately ,
95% of the time. Conversely, a sample with no real net -
activity above background will be reported as a positive value
, about 5% of the time."
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4 In addition, the licensee states in the ODCM that:
"LLD is predictive estimate (a priori) representing the capability of a measuring system, not after the fact (a posteriori) estimate of a particular sample. As such, typical values of E, V, Y and T should be used. Stated LLD's may not always be achievable due to background fluctuations, interfering radionuclides or other conditions affecting the normal measurement process."
This definition is consistent with the standard NRC definition presented in NUREG-0472.
On December 5, 1985, the Supervising Health Physicist discussed the LLNL results of the October 1985 sampling with Region V. The Cs-134 activity was reported at 8.6 E-9 uCi/ml; Cs-137 at 2.17 E-8 uCi/ml; Co-60 at 1.3 E-9 uCi/mi and Mn 54 at 5 E-10 uCi/ml. Although these activities were all less than the onsite LLD, a dose projection would place the 1985 exposure to the " hypothetical maximum individual" in close proximity to the T.S. limit if one assumed that activity was representative of the entire year's releases.
In a tel'ephone discussion between the Supervising Health Physicist and the Region V Chief, Facilities Radiological Protection Section, it was agreed that the licensee's Semiannual Effluent Release Report for July through December 1985 would address the LLNL water sample results, the LLD issue and the licensee's plans to submit the results of their evaluation.
The Semiannual Effluent Release Report was transmitted by letter I
(RJR 86-087) dated March 3, 1986, and contained the above information and a commitment to prepare and submit a special report by August 31, 1986, of the 1985 liquid radiological effluent release i source term.
In the course of preparing for a meeting to finalize the NRC l response to the 1984 liquid effluent issues (Paragraph 2 of this report) an NRC Licensing Project Manager became aware that the licensee may have changed their onsite LLD to facilitate the release of potentially contaminated liquid to the environment.
1 As a result of this information, Region V contacted individuals within the~11censee's organization by telephone on March 21, 27, and 28, 1986. Fron'these telephone discussions, Region V was informed
, of at least one instance when the routine three liter effluent water l
sample analyked according to the normal procedure of gamma counting
. for 2000 seconds showed identifiable and maasurable concentrations of cesium and the technician was directed by a management representative to recount the sample for 1000 seconds. Decreasing the counting time by a factor of two has the effect of reducing ,
the sensitivity considering all other parameters remain constant.
The 1000 second recount did not show any identifiable ~or measurable concentrations of cesium so the volume of liquid was released to the environment.
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- 5 Region V was told this matter had been brought to the attention of licensee management during December 1985 in the context of a ;
violation of T.S. and that management had concluded no violation had l occurred.
The purpose of this inspection was to address the following issues:
-Did the licensee change the onsite laboratory LLD to facilitate the release of potentially contaminated liquid to the environment?
-Did the release of radioactive material in liquid effluent during 1985 exceed the criteria in T.S. 3.17.27
-Has the licensee's management of liquid radioactive effluent been effective?
- 4. Lower Limit of Detection Technical Specification 4.21.1, " Liquid Effluents" Concentration, reads in part:
"The rad'ioactivity content of each batch of radioactive liquid waste to be discharged shall be determined prior to release by sampling and analysis in accordance with Table 4.21-1. The results of pre-release analyses shall be used with the calculational methods in the ODCM to assure that the concentration at the point of release is limited to the values in Specification 3.17.1.
" Post-release analyses of samples from batch releases shall be performed in accordance with Table 4.21-1. The results of the post-release analyses shall be used with the calculational methods in the ODCM to assure that the concentrations at the point of release are limited to the values in Specification 3.17.1."
- Table 4.21-1, Radioactive Liquid Waste Sampling and Analysis Program, l
requires that each batch of liquid waste to be released be sampled and l
analyzed for various radioisotopes. The minimum required LLD for mixed fission and activation products including Co-58, Co-60, Cs-134 and Cs-137 is stated as 5 E-7 uCi/al.
TheBasesofthiy,specificationreadsinpart:
"This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to area's beyond the site boundary will be less than the
- cUncentration levels specified in 10 CFR Part 20, Appendix B.
Table II. This limitation provides additional assurance that the l
1evels of radioactive materials in bodies of water outside the site will not result in exposures within: (1) the Section II.A Design ,
Objectives of Appendix I,10 CFR Part 50, to an individual, and (2) the limits of 10 CFR Part 20.106(e) to the population."
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The inspector noter that the word "not" has been included in the second sentence.
The same sentence from NUREG-0472, Revision 3. Standard Radiological Effluent Technical Specifications For Pressurized Water Reactors, reads:
"This limitation provides additional assurance that the levels of radicactive materials in bodies of water in UNRESTRICTED AREAS will 4
result in exposures within (1) the Section II.A design objectives of i Appendix I, 10 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the limits of 10 CFR Part 20.106(e) to the population." l The Regulatory Compliance Supervisor stated to the inspector that the word "not" had been deliberately inserted into the " Bases" during development of this specification because members of the licensee's organization recognized that the LLD values for their site might not be adequate to demonstrate compliance with the design objectives of Appendix I, Section iI.A.Section II.A. limits the dose due to liquid
- effluents to 3 mrem per year to the total body. T.S. 3.17.1, Dose, implementsSection II.A. of Appendix I.
The inspector found that'several individuals within the licensee's
- organization 'ere w not aware that the LLD values presented in Table 4.21-1
- were not intended to provide assurance that the T.S. 3.17.2 dose limits would be met. Specifically, the Radiation Protection Superintendent, Acting Chemistry Supervisor, and two Senior Chemistry and Radiatio'n Assistants all stated to the inspector that they believed if their onsite laboratory capability had an LLD of at least the value in Table 4.21-1 and they did not identify measurable radioisotopes in the liquid effluent releases, the dose limits of T.S. 3.17.2 and Appendix I would not be exceeded. All four CRPT interviewed confirmed that they had been told this was the case by their chemistry and radiation protection supervisors.
In a December 16, 1985, memorandum from the Supervising Health Physicist to the Manager, Nuclear Engineering, the Supervising Health Physicist presented the October 1985 LLNL sample results and described his awareness beginning in January 1985 that the onsite LLD's may not be adequate to assure compliance. The Memorandum described his efforts to
) evaluate the LLD issue, the lack of management support, his awareness of NRC interest, and proposed six specific actions to be accomplished. The issueofcommuni,gationsisdiscussedinParagraph,6ofthisreport.
10 CFR 50, Appendix I, Section IV.A. reads in part:
"A. If the quantity of radioactive material actually released in effluents to unrestricted areas frou a light-water-cooled nuclear power reactor during any calendar quarter is such that the resulting radiation exposure, calculated on the same basis as the respective design objective exposure, would exceed ,
one-half the design objective annual exposure derived pursuant to Sections II and III, the licensee shall:
"1. Make an investigation to identify the causes for such release rates:
8
7
- 2. Define and initiate a program of corrective actions:
and ...."
Paragraph B. adds:
"The licensee shall establish an appropriate surveillanc and monitoring program to:
"1. Provide data on quantities of radioactive material released in liquid and gaseous effluents to assure that the provisions of paragraph A of this section are met:"
As of April 1, 1986, since the Radiation Protection Superintendent apparently believed that the LLD values presented in T.S. Table 4.21-1 were adequate to meet Appendix I, no other appropriate monitoring program had been established to provide data on quantities of radioactive material released in liquid effluents to assure the dose criteria of Appendix I were met.
Failure to establish appropriate surveillance and monitoring procedures represents:an apparent violation of 10 CFR 50, Appendix I, (50-312/86-15-01).
The licensee's Semiannual Effluents Release Reports dated September 26, 1985, (RJR 85-491) and March 3, 1986, (RJR 86-087) presented in Table 2C the Rancho Seco onsite " Liquid Effluent-Lower Limit of Detection" for Cs-134 as less than 4.82 E-8 uCi/a1 and Cs-137 as less than 5.92 E-8 uCi/al. The licensee representative stated that the LLDs presented in Table 2C were based on a normal three litar liquid effluent sample counted for 2000 seconds using the average background counting rate on the gamma counting system.
i i The LLDs presented in this table were meant to show that the onsite capability clearly exceeded the values in T.S. Table 4.21-1 for the normal measurement procedure.
Since the licensee's counting system calculates an LLD for each measurement, the NRC Radiation Laboratory Specialist reviewed the licensee's methodology.
Rancho Seco uses " machine" generated "LLD" values to determine whether or not the LLD limits for specified nuclides are being met.- The procedure used for calculating-LLD is contained in the Canberra sof tware associated with the Canberra Spectran F gamma spectroscopy system. The document that addresses the software is Canberra Technical Reference Manual for Spectran F Version 2. This document, however, does not contain
. suffiefent information to determine exactly how the LLD is calculated.
Furthermore, several key equations in this document appear to contain typographical errors, and the discussions on LLD and LD (detection limit)
L seem to indicate that a procedure is being used to calculate LLD which is ,
j not consistent with the NRC definition of LLD.
In an attempt to resolve some of these issues, a telephone call was made to Mr. Markku Koskal of Canberra on Wednesday, April 2, 1986. On the a.
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basis of this discussion with Mr. Koskel, it appeared that appropriate LLDs were being generated. On the other hand, because a simple test involving a manual LLD calculation using raw spectral data would readily resolve the issue, it was requested that Rancho Seco perform manual LLD calc'ulations and compare these to the software generated LLD values.
This was done, and good agreement was obtained for the energy range tested. It can be concluded, therefore, that the software generated LLDs are consistent with the NRC definition of this term.
Accordingly, by having established confidence that the LLDs presented in computer printouts for liquid effluent analysis were credible, the inspectors reviewed records of liquid effluent releases made during June 1985 and the Chemical / Radiation Log for the first calendar half of 1985.
On March 20, 1985, an entry at 1730 in the Chemical / Radiation Log reads:
4 "Name deleted and name deleted concurred that we should count ARHUT for
, release for 1500 see to preclude obtaining a Cs peak which could prevent i
the RHUT'c release to the basin. 1500 see would meet CE and LLD on Canberra." Enclosure 4.1 " Rancho Seco Radioactive Liquid Waste Release Permit Regenerant Holdup Tank to Retention Basin" No. 85-76 indicates the "A" Regeneran.t Holdup Tank'(RHUT) containing 85950 gallons of liquid was released to the basin for discharge to the environment. The permit contains the comment "No Peaks" and lists the gross beta activity as 9.15 E-8 uCi/ml and H-3 as 1.42 E-5 uCi/ml.
Based on standard practice, all RHUT liquid release samples were normally counted for 2000 seconds at this time; however, the licensee only had a record associated with a 1501 second count at 1720 on March 20, 1985, of
! a three liter sample from A RHUT. This printout did not show any gamma t isotopes greater than LLD. Cs-134 had an LLD listed for this analysis of less than 8.23 E-8 and Cs-137 of less than 1.17 E-7 uCi/ml.
The Chemical / Radiation Log dated June 4, 1985, contains the following entry:
"85-98 B RHUT Scan Co-137 2.33 E-7 1 4.91 E-8 H-3 4 4.26 E-6 T Sean Repeat Cs-137 2.59 E-7 I Scan Repeat 1000 Sec Wo Peaks -
1638 B RHUT 85-98 H-3 4 4.34 E-6
! Gross S 2.93 E-7 t 4.04 E-8 l
. $ Scan-No Peaks (minor, but some peaks with.LLD's 'lP 5 E-7)"
Enclosure 4.1 for release 85-98 indicates 150,767 gallons were transferred from the B RHUT to the Retention Basin for release. The form ,
reads, "No i peaks" for mixed fission and activation products and H-3 i less than 4.26 E-6 uCi/ml. No entry is made regarding the gross beta activity.
. _,- . - - - . _.,_-m, . . _ _ . _ _ . - - , - , - . . . _ , . _ . , . . . , . . . _ , - - . , _ _ _ .-. ---_ - -
- o 9 The following table summarizes the results of gamma scans performed on three liter samples from the B RHUT tank on June 4, 1985.
Sample Time (PST) Counting Time Result (uC1/ml) 1020 2000 sec. Cs-134 < 1.2 E-7 Cs-137 2.72 E-7 1313 1000 sec. Cs-134 dC 1.25 E-7 Cs-137 < 1.67 E-7 1347 2000 sec. Cs-134 4 7.79 E-8 Cs-137 2.33 E-7 1701 1000 sec. Cs-134 4 1.51 E-7 Cs-137 4 1.88 E-7 In discussions with several chemistry and radiation technicians, the inspector was informed that when a 2000 second count showed identifiable peaks, the matter was brought to the SCRA's attention. The SCRA told them to recount the sample for 1000 seconds. If the 1000 second count did not show the presence of identifiable peaks and the LLD for this measurement as indicated on the printout was less than the 5 E-7 uCi/ml value in T.S. 4.21-1 then the release could be made. Some technicians
, stated that they did not believe this was the correct action, however, they did what they were told. Several individuals stated the motivation for the change in counting time stemmed from the excessive inventory of plant water and the licensee's public statement that they would not release any additional liquid radioactive effluents.
The Radiation Protection Superintendent and the SCRA told the inspector that their concern was the need to release water and that they believed that as long as the LLD for a given sample analysis was less than the l
T.S. number they would not exceed the Appendix I dose objectives. The Radiation Protection Superintendent stated that he did not become aware that the T.S. number (5 E-7 uCi/ml) might not be adequate to assure compliance with Appendix I until he received a capy of the Supervising Health Physicist " Draft LLD Study" on October 29, 1985.
The technicians recalled that the practice of recounts occurred on other occasions.
l The table below summarizes other examples noted in June 1985:
Release Initial Initial Final Final Counting Cs-137 Counting Cs-137 Date No, Tank Time Activity . Time Activity 6/6/85 85-99 ARHUT 2000 sec. 2.11E-7 1000 sec. 41.53E-7 uCi/ml 6/16/85 85-109 ARHUT No record 1700 sec. 41.03E-7 uCi/ml 6/17/85 85-110 BRHUT 2000 sec. 1.2E-7 1700 sec. 49.29E-8 uCi/ml Licensee procedure AP.306 V-13. " Lower Limit of Detection Count Time Determination," issued June 26, 1984, captures the essence of NRC's t
t 10 definition of LLD as presented in NUREG-0472, Revision 3, Table 4.11-1.
The procedure is designed to calculate the optimum counting time to meet the LLD minimum requirement specified in T.S. Table 4.21-1. The procedure does not indicate that the author realized that a far more sensitive LLD may be necessary to meet the dose limits of T.S. 3.17.2.
From discussions with the SCRA and review of data, it appears the first time AP.306 V-13 was fully implemented for detector 1 of the Canberra system was on July 30, 1985. At that time, a three liter liquid background sample was counted for 1000 seconds, five times. This test demonstrated that the LLD for Cs-134 was SE-8 uCi/ml and Cs-137 was 6E-8 uCi/ml. As a result of this test the licensee posted Enclosure 7.3,
" Liquid and Gaseous Effluent Release Recommended LLD Counting Time " on the Hot Laboratory bulletin board recommending a 1000 second count time for three liter effluent samples on Canberra detector 1.
Because technicians continued to believe a 2000 second count was more appropriate, during the remainder of 1985, of the 111 samples analyzed, 69 vere counted for 2000 seconds.
During the : initial phase. of the inspection the licensee was unable to locate the Canberra printouts for several of the initial 2000 second counts.
T.S. 6.10.2 reads: "The following records shall be retained for the duration of the Facility Operating License:...
"c. Records of gaseous and liquid radioactive material released to the environs."
During the subsequent inspection visits, the licensee was able to locate records except in two instances:
Release Permit No. Date Tank 85-203 10/29/85 ARHUT 85-213 11/13/85 ARHUT Failure to maintain records of liquid radioactive material released represents an example of failure to comply with T.S. 6.10.2 (50-312/86-15-02).
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Technical Specification 4.21'.1 reads in part: ~
"The radioactivity content of each batch of radioactive liquid waste to be discharged shall be determined prior to release by
. eampling and analysis in accordance with Table 4.21-1:
"Fcat-release analyses of samples from batch releases shall be performed in accordance with Table 4.21-1..." ,
a
s 11 T.S. Table 4.21-1 footnote c. reads:
"Other peaks which are measurable and identifiable, together with the listed nuclides, shall also be identified and reported.
Nuclides which are below the LLD for the analysis should not be reported as being present at the LLD level.
T.S. 6.9.2.3, Semiannual Radioactive Effluent Release Report, reads in rsrt 6.9.2.3.1:
"The radioactive effluent release reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21,
' Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants,' with data summarized on a quarterly basis, following the format of Appendix B thereof."
Regulatory Guide 1.21 reads in Paragraph B.2:
"In many, cases the crit'eria for sensitivity of effluent measurements have been modified to reflect as low as practicable dose considerations in the offsite environs; i.e., the sensitivity of effluent measurements should be sufficient to detect concentrations which, when dispersed in the offsite environs, would result in a dose to individuals of a small fraction of natural background radiation."
and Paragraph C.10 "The sensitivity limits given for radioactivity analyses in Appendix A of this guide are based on the potential significance in the environment of the quantities of radioactive materials released.
For some radionuclides, lower detection limits than those given herein may be readily achievable and when measurements below the stated sensitivity limits are attained, the results should be recorded and reported."
The licensee's Semiannual Radioactive Effluent Release Report for the first calendar half of 1985, dated September 26, 1985, stated in Section B. Paragraph C. -
" LIQUID EFFLUENTS "As a result'of steam generator tube failures in May 1981. November
- 1982, September 1983, and July, August, and September 1984, a significant quantity of radioactive primary fluid has been circulated through the steam generation cycle. After the September 1983 occurrence, a small leakage path appeared to remain, on the .
order of 0.07 spa, which could not be located even after extensive
! investigating and testing.
" Residual gamma emitters from the secondary system have not been released in the waste water stream during this report period.
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. "There were 199 batch releases from Regenerant Holdup Tanks with the material ultimately released by 46 Retention Basin discharges.
Liquid releases are summarized in Table 2A and the isotopic contents are detailed in Table 2B."
Table 2A contained the following statement for fission and activation products:
"N/A - no releases containing detectable fission or activation products were made in the period of January through June 1985" On June 6, 1985, the Radiation Protection Superintendent called the NRC Region V Chief Facilities Radiological Protection Section, to discuss footnote c. The licensee documented the call as follows:
" Reason for Call:
Resolve meaning or interpretation of second sentence Table Notation 'C' Table 4.21-1 page 4-71.
" Resolution Reached:
If's nuclide is below minimally required LLD (gj ' uCi/cc) but is a positive value it must be recorded and repo< ed."
Failure to report positive results for Cs-137 activity which was
, identified and measured on Jane 4, 6 and 17, 1985, in liquid effluent releases 85-98, 85-99 and 85-110 in the Semiannual Radioactive Effluent Release Report dated September 26, 1985, is considered an apparent violation of T.S. 4.21 (50-312/86-15-03).
T.S. Table 4.21-1 requires that a monthly composite be collected from each Batch Waste Release Tank for quarterly analysis of Sr-89 and Sr-90.
From January to November 1985 the licensee interpreted the requirement such that samples were only taken from releases which showed gross beta, gamma or tritium activity in excess of their respective LLDs. This interpretation is considered to be inconsistent with the NRC
- intrepretation of the T.S. (50-312/86-15-04).
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- In November 1985, after distribution of the " Draft LLD Study," composite samples were collected from all batches of liquid released from the RHUTs to the basins.
As previously dfscussed in Paragraph 3 of this report, the licensee representative was aware on November 26, 1985, that the composite samples could be analyzed for gamma emitting isotopes to aid in better determining the 1985 liquid radioactive effluent release source term.
A December 31, 1985. Memorandum from the Manager, Nuclear Engineering, to tha Radiation Protection Superintendent requested that the composite samples be analyzid by their contractor for radioisotopes of cesium. .
On April 1, 1986, the inspector inquired as to the results of the analyses. The Radiation Protection Superintendent stated that they had
j
^ . 13 received the results but they were not prepared to accept the data provided by CEP.
l The inspector was allowed to review a letter dated February 24, 1986, I from CEP to the licensee representative. The results transmitted l indicated unrealistically high concentrations during the months of February, March and April; insufficient volumes of liquid to make the !
measurement for May, August and September; November and December samples had not yet arrived; January was below their detection limit; and June, July and October showed measured concentrations well above CEP's LLD for Cs-134 and Cs-137.
The licensee had decided that the high activities observed for February, March and April were the result of using contaminated glassware. No other explanations were offered regarding the remaining months and no one had initiated dose calculations.co determine compliance with T.S. 3.17.2 as required by T.S. 4.21.2 Doses.
T.S. 4.21.2 reads:
" Dose Calculations "Cumulaiive dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calculation Manual (ODCM) at least monthly."
ODCM Section 2.3, Compliance With 10 CFR 50 Liquid Radioactive Effluents, reads in part:
"It is necessary to demonstrate compliance with 10 CFR 50 Appendix I only if liquid effluents contain measurable quantities of radionuclides. The point of liquid effluent radionuclide
! quantification is defined as the regenerate holdup tanks. The liquid effluent is to be analyzed in accordance with Technical Specifications 4.21.1."
As of April 1, 1986, the licensee had maintained that since they had not measured Cs-134 and Cs-137 in liquid effluent, they were not required to perform the dose calculations for these isotopes. On April 2, 1986, the inspector requested that the licensee expeditiously resolve their 1985 liquid effluent source term, inform Region V of their conclusions, perform the required dose ~ calculations, and submit the required reports if the results ihdicate the limits of T.S. 3.17.2'had been exceeded.
At the conclusion of Paragraph 3, the following question was presented:
. "- Did the licensee change the onsite laboratory LLD to facilitate the release of potentially contaminated liquid to the environment?"
This paragraph documents that the Rancho Seco onsite organization altered ,
the counting times of liquid affluent samples to facilitate the release of liquid as necessary to relieve operational restraints. The individuals involved stated that they believed that as long as the concentration was less than SE-7 uCi/ml, the design objectives of 10 CFR 50 Appendix I and dose limits of T.S. 3.17.2 would not be exceeded.
_ . _ , . . - _ - _ . _ , - - . . _ _ . . - , - . - - . - . . .- _ , . . . _ . - ~ -- _ , - - , .
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- 5. Compliance With Liquid Effluent Dose Objectives 10 CFR 50 Section 50.36a contains provisions designed to assure that releases of radioactive material from nuclear power reactors to unrestricted areas during normal reactor operations, including expected operational occurrences, are kept as low as practicable. In July 1984, l T.S. 3.17.2 became effective. This specification reads: I l
"The dose or dose commitment to a member of the public from radioactive materials in liquid effluents released beyond the site i
, boundary shall be limited:
"a. During any calendar quarter to 1.5 mrem to the total body and to 5 mrem to any organ; and "b. During any calendar year to 3 area to the total body and to 10 mrem to any organ."
The action statement requires that:
"With.the calculated dose or dose commitment from the release of radioactive materials in liquid effluents exceeding any of the above
- limits, prepare and submit to the Commission within 30 days a Special Report. This Report will identify the cause(s) for exceeding the limit and define the corrective actions to be taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits."
As a result of a licensee commitaant on Apri'. 2,1986, to provide their position with respect to the activity reported by CEP from the analysis of composite samples, the licensee submitted a letter to Region V on April 17, 1986 (RJR 86-135).
The April 17, 1986, letter contained three enclosures and five immediate actions to preclude noncompliance with the 10 CFR 50 Appendix I design objectives in 1986.
Enclosure I - A letter dated April 8, 1986, from CEP to the licensee summarizing the composite sample results for 1985.
Enclosure II - An arithmetic composite of LLDs for 1985.
Enclosure III - A summary of liquid waste released, and total volume released during 1985.
. The immediate actions included:
"1) All liquid samples for effluent release are being counted for 2000 seconds; 2) the average plant effluent release rate has been .
increased to 5000 gym to more closely represent'a 'non-dry site' power station; 3) all documentation relating to liquid effluent releases are placed in a separate folder which will contain all the paperwork associated with the release (i.e., gamma scans, beta results, tritium results, chemical data sheets (Enclosure 4.2) and (Enclosure 4.1) of Administrative Procedure (AP) 305 13; 4) the e
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stated LLD values from the gamma scan, gross beta, or tritium analyses will be written on Enclosure 4.1 of AP 305-13; 5) a change is in progress to clarify compositing requirements and require compositing be performed in the secondary lab with only clean glassware to preclude contamination of samples." ;
l Review of Enclosure I indicated that four monthly composites had positive I results for cesium isotopes, three months showed no detectable activity and five months were not of use due to either not enough sample or contaminated glassware. Based on the licensee's evaluation of the four months of clearly indicated cesium activity, they initiated development of the 30-day Special Report required by T.S. 3.17.2.
4 The inspector noted that Enclosure I did not contain: a result for that composite classified as a "non-radioactive" release volume during December 1985; results of alpha,.Sr-89 and Sr-90; and an explanation of why sample results were not available for May, August, and September.
The licensee responded by a memorandum on April 21, 1986, which indicated l the December "non-radioactive" releases contained 2.8E-8 uCi/ml 2 8E-9 uCi/ml of Cs-137. The alpha, Sr-89, and Sr-90 activities were all less than LLD, and sample volumes for May, August, and September were not available due to repeat analyses for gross alpha and strontium during l those months.'
Based on an inoffice review of the potential liquid radioactive release !
source term and the licensee's September 27, 1984, commitments to reduce liquid effluents, the inspector concluded that an additional site visit
, would be appropriate to determine the origin of released activity and the potential that T.S. 3.17.2 might have been exceeded.
, On April 29, 1986, the inspector returned to the site and corporate office. This visit found that since 1983 the licensee has engaged in a water management practice inconsistent with the description in the Final Safety Analysis Report (FSAR).
- 10 CFR 50, Appendix A, 1.5.51 CRITERION 60 - CONTROL OF RELEASES OF RADIOACTIVE MATERIALS E THE ENVIRONMENT re. ads:
"The nuclear power unit design shall include means to control suitably the telease of radioactive materials in gaseous and liquid effluents and to handle radioactive solid wastes produced during
, normal reactor operations, including anticipated operational occurrencert Sufficient holdup capacity shall be provided for retention of gaseous and liquid effluents containing radioactive materials, particularly where unfavorable site environmental conditions can be expected to impose unusual operation limitations
. upon the release of such effluents to the. environment."
Section 1.5.51 of the FSAR reads:
"The radioactive waste system collects, segregates, processes, and disposes of radioactive solids, liquids, and gases in such a manner that suitable control is provided over releases in order that l
1 l
\ .
16 numerical guidelines can be met for as low as practicable as defined in 10 CFR 50, Appendix I."
The updated FSAR submitted July 22, 1982, and subsequent amendments through July 1985, provide information in Section II., Radioactive Waste and Radiation Protection ,that:
"The radioactive waste disposal systems provide for the controlled handling and disposal of liquid, gaseous, and solid wastes. The systems are designed to ensure that plant personnel and the general public are protected against excessive exposure to radiation from wasces, in accord with 1Laits defined in 10 CFR 20, 10 CFR 50, and 40 CFR 190. i "The systems minimize or preclude discharge of radioactive liquids, gases, and solids of station origin to the surrounding environment.
Liquids are not discharged to the environment during normal operation but are processed and held for reuse or for solidification and shipment offsite by an NRC-licensed contractor."
^
The licensee has, as a result of limited radioactive water storage capacity, rou.tinely transferred water from the Demineralized Reactor Coolant Storage Tank (T-621) to the RHUTs (T-950 A and B) for discharge to the environment. >
In response to a reque:t by the inspector, the licensee determined that r during 1985 787,500 gallons were transferred from T-621 to the RHUTs and released to the environment. Management issues surrounding this transfer are discussed in Paragraph 6 of this report.
3 The Demineralized Reactor Coolant Storage Tank is a 450,000 gallon,
- quality class 1, seismic, category 1 tank which receives water from the coolant radwaste system. The licensee representative stated that due to chronic steam generator tube leaks and plant operational configurations, water had to be transferred to the RHUTs.
No specific gamma activity analyses were made each time tha transfers took place. The licensee was concerned that the tritium concentration could be limiting; therefore, the tank was sampled for tritium concentration 26 times in 1985. The average tritium activity was 2.46 E-2 uCi/al. The inspector was told that the standard practice was to use ,
a temporary piping system to pump between 10,000 to 40,000 gallons to the RHUT as a functi6d'of tritium activity. The RHUT would then be filled from the normal secondary system sources or the service water system to dilute the tritium as necessary to assure that the concentration limits of 10 CFR 20 were'not exceeded.
Based on the limited sample data available, the inspector prepared a ,
"best" estimate of activity released during 1985. The inspector used four gamma scan results made available by the licensee to prepare the .
following source term estimate of cesium activity. It must be noted that other isotopes including iodine-131, antimony 124 and 125, silver 110M, niobium 97, and cobalt-58 were observed in low concentrations in some samples. The inspector selected the highest cesium isotopic activity 4
y?
l
' . 17 from either the CEP composite for that month or the T-621 sample as the best estimate. This appears' justified since at times additional sources of activity could have been added to the RHUT resulting in the CEP composite calculated activity exceeding the T-621 source term. For example, in December 1985, the 5000 gallons of water from T-621 transferred to the RHUT for discharge contained 37 uCi of Cs-137, but the activity calculated from composite samples of the 1,560,000 gallons of "non-radioactive" RHUT discharges using.the measured activity of 2.8 E-8 uCi/ml amounted to 165 uCi of Cs-137.
The data below summarizes the 1985 "best" estimate cesium source term:
- Best Estimate (uCi)
Month CEP Activity T-621 Activity Cs-134 Cs-137 January Not measured .Not measured February Contaminated samples Not measured March Cs-137 - 713 uCi Cs-137 - 267 uCi 713 April Contaminated samples No water transferred May No measurement made No water transferred June Cs-134'= 176, = 511 C's-137 = 171 176 511 July .Cs-134.=,302, Cs-137 = 465 No data available 302 465 August No measurement made Cs-134 = 177, Cs-137 - 225 177 225 September No measurement made Cs-134 = 1137,Cs-137 - 1443 1137 1443 October Cs-137 - 153 No duca available 153 November Not measured No data available December Cs-137 - 165 , Cs-134 = 29, Cs-137 - 37 29 165 TOTAL 1821 3678 Note -The cesium' estimate only includes releases made during the six months of 1985 for which there were sample data that indicated activity greater than the cesium LLD.
-Large volumes of water released from the PEUTs classified as
. ' con-radioactive" from January through October 1985 were not composited for gamma isotopic analysis.
-Other games isotopes were not considered to simplify the presentation.-
In order to establish the credibility of this estimate, the inspector compared the 1985 tritium activity released as reported by the licensee in their Semiannual Radioactive Effluent Release Reports to the tritium released from T-621 to the REUT for discharge using the average activity
. from the 26 tritium samples.
The licenses had reported 89.86 curies had been released. The inspector calculated 73.23 curies originated from T-621. From this, the inspector ,
concluded that the "best" estimated cesium activity most probably l underestimates the actual release source term.
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Technical Specification 4.21-2, Doses, reads:
. " Dose Calculations
" Cumulative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calculation Manual (ODCM) at least monthly."
The Bases reads in part: ,
"The Dose Calculations Methodology in the ODCM implements ths requirements in Section III.A of Appendix I that conformance with the guides of Appendix I is to be shown by calculational procedures based on models and data such that the actual exposure of an 4
individual through appropriate pathways is unlikely to be substantially underestimated."
i From Revision 3, effective September 23, 1985, of the licensee's ODCM, the calculated dose from liquid effluents released during 1985 using the nonconservative best estimate of activity is 3.89 area to the total body
- of the hypothetical maximally exposed member of the public.
At the conclu'sion of Paragraph 3, the following question was presente1:
"- Did the release of radioactive material in liquid effluent during 1985 exceed the criteria in T.S. 3.17.27" This paragraph documented that radioactive material in liquid effluents exceeded the criteria in T.S. 3.17.2.
2 Although this appears to represent an apparent violation of T.S. 3.17.2 b (50-312/86-15-05), it is reasonable to expect that no real member of the i public actually received a dose greater than this value as a result of the 11(uid releases made during 1985.
In the course of developing the source term, the inspector found several additional deficiencies including arroneous data in the ifcensee's Semiannual Radioactive Effluent Release Reports, failure to complete the
~
land use census required by T.S. 4.27 failure to revise the ODCM consistent with T.S. 6.16, failure to follow procedures required by
, T.S. 6.8, failure to perform safety evaluations required by 10 CFR 50.59, and failure to update the Final Safety Analysis Report as required by 10 CPR 50.71(e) '-'These findings involve management issues which are described in.the next paragraph.
- 6. Management Issues
- A. Changes As previously noted in the introduction to Paragraph 5 of this ,
report, the action statement associated with exceeding the dose limit of T.S. 3.17.2 recognizes the limited safety significance of I the Appendix I dose values and requires a special report that:
--. ,.....---,-,-r--.c.,_,,_,, ---- - - - ,,- -- ~-, --,-,---v,-,. - - - -- , - - - - - - s------,- - ,e ~ - e---~" ~ ~ ~ ~ , - ~ - * - ~ - " ^ - - ~ ' ~ - - ' -
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19 i *
"will identify the;cause(s) for exceeding the limit and define the corrective actions to be taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits."
The licensee's Special Report No. 84-07 submitted in response to this requirement en September 27, 1984, described the cause for exceeding the dose limits for 1984 as a small but continuous leak in the "B" Once Through Steam Generator. The report reads:
"The path that the radioactive material takes to get from the secondary system to the general public is as follows.
Backflush water, regenerant waste and flush water from th polishing domineralizera flow to the Polishing Domineralizer
- Sump (PDS). Also, leakage from the secondary system generally flows to the Condensate Pit Sump where it is transferred either to the PDS or directly to one of the Regenerant Hold Up Tanks j (RHUTs). The PDS is typically pumped to one of the RHUTs,
! which, when full, are agitated, neutralized, and sampled. The results of the. sample are used to determine the total release activity for each isotope. Based on this data, a Liquid Waste
! Release Permit is generated, then the tank is pumped to one of the Retention Basins. When the Basin is full, it is recirculated ard sampled to determine a dilution rate that l would conform to appropriate limits. Samples are also taken during the discharge to provide assurance that regulations are met. Previously 10 CFR 20, Appendix B requirements were applied as limits as the discharge left the site. The District i now is limiting its discharges so that 10 CFR 50 Appendix I limits will not be exceeded."
l l
Based on this inspection, it appears that the cause and pathway were not entirely correct. Specifically, beginning in 1983, the licensee initisted a procedure which allowed the frequent transfer of water
~
recovered from the liquid radioactive vaste treatment systems to the RHUTs for release to the environment.
10 CFR Part 50.59(a)(1) reads:
"The holder of a. license authorfzing operation of a production or utilisation facility may (1) make changes in the facility as described in the safety analysis report,' (ii) make changes in the procedures as described in the safety analysin report, and (iii) conduct tests or experiments not described in the safety analysis report, without prior Commission approval, unless the proposed change, test or experiment involves a change in the technical specifications incorporated in the license or an unreviewed safety question."
The updated FSAR submitted July 22, 1982, and subsequent amendments I through July 1985, provide the following information in Section 11.,
Radioactive Waste and Radiation Protection, that:
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"The liquid waste. systems are designed to permit plant operation without discharging radioactive liquids to the environment under normal operating conditions. The boric acid concentrator and miscellaneous waste evaporator can each process waste liquids in excess of the maximum expected waste generation rates. The coolant vaste receiver and holdup tanks
. are sized to store one reactor coolant system volume of waste during an evaporator outage or during maintenance.
, "The coolant waste system ir a closed loop water system with I
the recovered water and boric acid stored onsite for reuse.
"The miscellaneous liquid radwate system, through the use of the miscellaneous water holdup tank and the shipment of concentrated wastes offsite by an NRC-licensed contractor, allows normal operation without requiring the discharge of liquids from the system. The entire liquid waste processing
, system is contained within the Auxiliary Building. Therefore, any leaks will be retained within the building, collected in the sumps, and reprocessed through the miscellaneous liquids j radwaste system.
1 -
"All vents, drains, and secondary flow paths in the liquid radwaste system are shown in Figures 11.1-4 and 11.1-5. The system is designed so that no liquid radwaste will be released
- to the environment."
Section 11.1.2.2.2., Miscellaneous Liquid Radwaste System, reads:
"In addition, spent regenerant wastes from the polishing domineralizers can be processed if they contain radioactivity
- as the result of operation with a small steam generator tube
- leak."
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Based on review of licensee records, it appears that on December 7, j 1982, a temporary change to Procedure A.29, " Waste Water Disposal l System," was implemented which allowed radioactive water to be pumped from the Domineralized Reactor Coolant Storage Tank (T-621) through a temporary conduit to either Regeneration Hold-up Tank (T-950 A or B) for ultimate release to the environment and the Principle Reguistory Compliance Engineer was unable to provide any indication that an evaluation had been performed to determine if a change in the T:S. was required or if an unreviewed safety question was involved. The temporary change expired on January 7, 1983, and l was reestablished on February 8, 1983, and then expired on March 30 I 1983.
Again, on January 6, 1986, a temporary change to Procedure A.10
" Demineralized Reactor Coolant Storage System," was implemented which allowed radioactive water to be pumped from T-621 through a ,
plastic pipe to either T-950 A or B for ultimate release offsits and the Principle Regulatory Compliance Engineer was unable to provide any indication that an evaluation had been performed to determine if a change to T.S. was required or an unreviewed safety question was involved.
E
- - * . 21 From January 1983 through March 13, 1986, the licensee routinely transferred liquid through various conduits including firehose and plastic pipe from T-621 located within the tank farm to either T-950 A or B which are located in an uncontrolled area such that failure of the temporary conduit might have resulted in an uncontrolled
, release of radioactive material to the surface waters.
On May 15, 1986, the inspector physically observed that the plastic pipe which had been connected to T-621 drain line had been removed leaving the exposed open pipe in close proximity to the tank. The licensee representative stated that the temporary pump which had been installed in the system had a flow rate of 166 gallons per minute.
T.S. 3.17.3, Liquid Holdup Tanks, limits the quantity of radioactive I
material which can be contained in the RHUTs and outside temporary storage tanks to 10 Curies. T.S. 4.21-3 contains the following comment:
" Tanks included in this specification are those outdoor tanks that are not surrounded by liners, oikes, or walla capable of holding the tank' contents and that'do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system."
The connection of a non-quality class temporary piping system, with no automatic isolation capability, to T-621 raised the question as to whether the licensee had performed the weekly surveillance on T-621 to determine that the activity was less than the 10 Curie limit while the temporary system was in operation. The licensee indicated that the surveillance had not been performed. The inspector attempted to review the accident analysis for failure of either T-621 or the Borated Water Storage Tank (450,000 gallon) since both are outdoors, not surrounded by liners, dikes, or walls capable of holding their contents, and the licensee does not perform
! the weekly surveillance. Neither the FSAR nor the licensee presented a safety analysis which would bound these tank failures.
This matter has been brought to the attention of NRR (50-312/86-15-06).
I The installation of a piping system specifically intended to j transfer vs,ter from the liquid radioactive treatment system to the
! RHUTs for r4Isade to the environment without first performing a
! safety evaluation is considered an apparent violation of 10 CFR 4
50.59 (50-31,2/86-15-07).
10 CFR 50.71(a) requires in part that each person licensed to 1 operate a nuclear power reactor shall annually update the final I safety analysis report (FSAR) to assure that the information included in the FSAR contains the latest material developed. The .
update must be submitted to the NRC and shall contain all the changes necessary to reflect information and analyses submitted to the Commission by the licensee or prepared by the licensee pursuant to Commission requirement since the submission of the original FSAR or, as appropriate, the last updated FSAR.
i .
- . 22 Based on discussions with licensee representatives and review of records, including Control Room Logs, it appears that the licensee has discharged liquid radioactive effluents from T-651 to the RHUTs for release to the environment from early 1983 through March 13,
- 1986 and did not update the FSAR to reflect this information. This represents an example of failure to comply with 10 CFR 50.71(e)
[ (50-312/86-15-08).
! Failure to perform the safety evaluations and update the FSAR is considered an example of failure to properly manage changes at the facility.
B. Procedures
- 1. T.S. 6.8, " Procedures," reads in part that, " Written procedures shall be established, implemented and maintained covering the j activities referenced below: a. The applicable procedures 1 recommended in Appendix 'A' of Regulatory Guide 1.33, November i 1972." Regulatory Guide 1.33, November 1972, recommends in G.,
" Procedure for Control of Radioactivity (For Limiting Materials Released to Environment and Limiting Personnel Exposure)." that
, procedures be developed for liquid radioactive waste systems including discharging of effluents.
Based on discussions with licensee representatives and review of records, it appears that from March 30, 1983, to January 6, 1986, no procedure was maintained which controlled the transfer of radioactively contaminated water from the Domineralized Reactor Coolant Storage Tank (T-621) to the Regenerate Hold-Up Tanks (T-950 A and B) for ultimate release to the environment.
4 During 1985, about 787,500 gallons were transferred from T-621 to T-950 A and B and released to the environment.
1 Based on review of the Control Roon Logs for March 1986 and document control records, it appears that on March 6, 1986, the temporary change to Procedure A.10. " Demineralized Reactor Coolant Storage System," which authorized transfer of water from T-621 to T-950 A and B was not maintained in that the procedure expired and a transfer of 6,000 gallons was made to T-950 A on March 10, 1986, and 15,000 gallons were transferred to T-950 B on March 13, 1986.
- 2. T.S. F:8.3 reads: " Temporary changes to procedures of 6.8.1
~
above may be made provided:
"a. The intent of the original procedure is not altered.
"b. The change is approved by two members of the plant i management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected. .
"c. The change is documented, reviewed by the PRC and approved by the Plant Superintendent within seven (7) days of implementation."
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l Based on review of the Procedural Change Approval Form and discussions with the Principle Regulatory Compliance Engineer, it appears that on January 6, 1986, a temporary change to Procedure A.10 " Demineralized Reactor Coolant Storage System,"
was approved and implemented which allowed pumping water from T-621 to T-950 A and B for offsite release without review by the Plant Review Committee (PRC). From January 6, 1986, to March 6, 1986, the licensee estimates that about 350,000 gallons of water were transferred.
In addition, the inspector noted that AP.2 Revision 21. Review, Approval and Maintenance of Procedures, had not been developed consistent with this T.S. in that it does not require temporary changes to be reviewed by the PRC. The Principle Regulatory Compliance Engineer informed the inspector on May 21, 1986, that this issue had be'en previously addressed by the PRC and that they believed the previous NRC Senior Resident had agreed that the review of non intent changes to procedures could be delegated to a Group Supervisor, reviewed by the PRC Chairman and approved by the Plant Superintendent. The inspector i
commented that if T.S. 6.8.3.c. were revised, their technique would be considered acceptable. In any case, the inspector i
considered the revision to A.10 to be an intent change in view i of the FSAR information.
Failure to implement and maintain procedures is considered an apparent violation of T.S. 6.8 (50-312/86-15-09).
The establishment, implementation, and maintenance of procedures is a management function. It is the inspector's I
conclusion that the proper establishment of these procedures considering the guidance provided in IE Circular No. 80-10:
10 CFR 50.59, " Safety Evaluations for Changes to Radioactive Waste Treatment Systems," could have resulted in recognition of the need to perform a 50.59 review, update the FSAR, and assure proper sampling of T-621 prior to transfer such that compliance with T.S. 3.17.2 could have been achieved.
C. Quality of Technical Work and Reports
, 1. The licensee's reports involving liquid radioactive effluents have (gequently contained inaccurate information and have not been submitted in a timely manner.
- a. Semiannual Radioactive Effluent Release Report, dated
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September 26, 1985, was required pursuant to T.S. 6.9.2.3 to be submitted within 60 days after July 1, 1985.
The statement in Table 2A that "no releases containing detectable fission or activation products -
were made in the period of July through June 1985" is incorrect as previously described in Paragraph 3 of this report.
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Table 2C, Liquid Effluents Lower Limit of Detection,
! may be in error since the licensee changed the ,
1 laboratory capability by altering the sample counting i time.
Section H., Estimation of Error, presents an inaccurate evaluation of the error associated with the reported releases. Since this same data was presented again in the second half 1985 report, the Radiation Laboratory Specialist performed the following review of the licensee's error analysis for liquid effluents.
Reportedly, for liquid releases, the error analysis includes error contributions due to sampling, volume
! measurements, and counting statistics.
The error formula Rancho Sect usesa f fi"I " *"E activation products is 1 (4* +10)ff2 (page 26).
, The 10 term is apparently the 10 (%) error they list for volume of' water. This leaves CP as due to a
, tombined sampling error and counting statistics tern; or either sampling error or counting statistics alone I with the other term being zero. This does not calgulate to 2%, the value given on Page 27, even if Cr* is zero.
For lov level samples with concentrations near the LLD, the counting error term would normally be 5-10 percent. Sampling error would probably also be in
- this range.
I 2 2 2
( (7.5) + (7.5) + (10) )1/2 =
14.6 (%)
For 95% confidence interval (21r"), the 14.6% value has to be multiplied by 2.
The licensee intends to revalidate their entire error l analysis. The licensee's corrective action regarding j this matter will be reviewed in a subsequent inspection (50-312/86-15-10).
~~
The licensee included a copy of Revison 3 of the ODCM. Since Revison 3 had an' effective date of September 23, 1985, its adequacy will be addressed with the evaluation of the second half 1985 report.
l The inspector observed that T.S. 6.9.2.3.1 incorrectly refers to T.S. 6.14 in describing what information must be included with revisions of the .
ODCH. The licensee was encouraged to correct this error (50-312/86-15-11).
Section I Table-1, Page 29, items 11 and 12 appear to be in error.
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- b. Semiannual Radioactive Effluent Release Report, dated March 3, 1986, contained the following errors.
"The estimates of radiat. ion dose equivalent to the non-occupational maximally exposed individuals are one or more orders of magnitude smaller than the limits of 10 CFR Part 50, Appendix I."
By virtue of the December 16, 1985, memorandum from the Supervising Health Physicist to the Manager Nuclear Engineering; the fact that effluent counting minipulation had been brought to the attention of all levels of the facility management including the Assistant General Manager Nuclear in December 1985 and January 1986; and that CEP data had been received which raised obvious questions if Appendix I had been met, the inspector considers this statement to be misleading.
The report reads on Page 7:
"The only gaseous abnormal release was associated with a reactor transient on December 26, 1985.
Radioactivity was released via primary-to-secondary i leakage hence to atmosphere from secondary safety relief and dump valves."
i This statement is incorrect. Nearly all of the 32.7 Curies released originated from the make-up pump
, failure and were discharged via the plant vent stack.
Table III-C contains the same LLD values presented in the previous report.
Table IV-A, Waste Disposal Summary, incorrectly reports the solid radioactive waste data for the period July through December 1985. The Supervising Health Physicists stated the data reported is for the entire year.
'The Estimation of Error is again incorrect.
~~
Table VI-B, Page 31, is in error. It appears to be a reprint of the data contained in Section I. Table 2, Page 30, of the previous report.
Section VIII, Page 123 reads:
"No changes were made to the ODCM during this period." This is incorrect. Revision 3 to the .
ODCM became effective September 23, 1985.
On April 29, 1986, the Assistant General Manager agreed that the reports needed to be corrected (50-312/86-15-12).
I
1e . 26
- 2. T.S. 6.9.2.2, Annual Radiological Environmental Operating Report, reads in part that: " Routine radiological environmental operating reports covering the operation of the unit during ,
6 the previous calendar year shall be submitted prior to May 1 of each year."
Based on discussions with the Supervising Health Physicist on April 29, 1986, the inspector learned that the annual land use census required pursuant to T.S. 4.27 had not been completed due to ongoing litigation with the near site residents. Since the T.S. requires the results to be included in the Annual Radiological Environmental Operating Report, it did not seem likely that the report could be submitted on time.
On April 30, 1986, the licensee advised that the required report would be submitted by May 30, 1986.
- 3. T.S. 6.16, Offsite Dose Calculational Manual (ODCM) reads in 6.16.2: "Any changes to the ODCM shall be made as follows:
"A. Licensee-initiated changes:
"1. Shall be submitted to the Commission by inclusion in the Semiannual Radioactive Effluent Release Report and shall contain:
"a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should
- consist of a package of those pages of the ODCM to be changed with each page numbered and
! provided with an approval and date box, together with appropriate analyses or evaluations
, justifying the change.
"b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and..."
Revisfod '3*of the ODCM effective September 23, 1985, was supplied with the Semiannual Radioactive Effluent Release Report dated September 26, 1985. That report and the subsequ'ent report did not contain information to totally
- ' support the rationale for the change. In addition, the change included a revision of the bioaccumulation factor for cesium from 2000 pCi/kg per pCi/1 to 1500 pCi/kg per pCi/1 without a
- determination that the change will not reduce the accuracy of .
the dose determination.
Failure to provide the required supporting data represents noncompliance with T.S. 6.16 (50-312/86-15-13).
4 ,, ,
- 27 The Supervising Health Physicist indicated to the inspector that he had not made himself familiar with this section of the T.S.
- 4. On May 9, 1986, the licensee issued Special Report No. 86-08, Preliminary Calculated Dose to the Public Excee gng the Numerical Design Objectives o_ff 10 CFR 50, g et. dix I.
This report stated a preliminary evaluation found that the
- quarterly numerical design objectives may have been exceeded for the third quarter of 1985 and the annual design objectives
. may have been exceeded for 1985. The licensee stated the final
- report would be submitted in 30 days.
The timeliness, accuracy and completeness of technical work in the radiological effluent management area indicates inadequate staffing, training, peer review and management oversight.
D. Communications Action VII of the licensee's September 27, 1984 Special Report reads:
"The District has initiated a policy that all releases will be controlled such that Technical Specification 3.17.2 limits will 4 not be exceeded. All sampling of the RHUTs and releases of liquids will be based on this objective. The Chemistry and Radiation Protection personnel responsible for evaluating the releases have been instructed concerning these objectives.
This Action, coupled with Action IX will provide a second level j of control beyond the other near term actions specified herein.
i Status: Implemented."
. During this inspection, when presented with this licensee l commitment, the Supervising Health Physicist, Radiation Protection
, Superintendent, Regulatory Compliance Supervisor. Acting Chemistry ,
Supervisor, and SCRAs all stated that they had never seen the
- commitment. The chemistry and radiation protection personnel stated that the only direction they had received was that they were not to 4
discharge any radioactive material other than, tritium in liquid affluent releases. They understood this direction was satisfied if
, they did not detect any gamma isotopic activity in excess of SE-7 uCi/ml in the RHUT releases.
Whe5'the inspector brought this information to the attention of the Assistant Ceneral Manager, Nuclear (AGMN), the AGKN expressed frustation and indicated that he would look into the matter.
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Three other points were noted:
i
! - The Regulatory Compliance Supervisor indicated that he and l others involved in implementing the Appendix I T.S. changes were aware that the LLDs were not adequate to assure compliance t
e 4
,.w.,-
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(, s. A
. 28 with the dose limits of T.S. 3.17.2. This information apparently was not provided to the 1985 operating chemistry and radiation protection personnel responsible for evaluating the releases.
The Supervising Hecith Physicist was aware of the LLD/ Appendix I issue in early 1985 but again this concern was not translated into action.
Technicians expressed their concera to their supervisors that the adjustment of counting time obscured the presence of radioactive material. Again, an opportunity to resolve the issue in a more favorable manner was not realized.
At the conclusion of Paragraph 3, the following question was presented:
" Has the licensee's management of liquid radioactive effluents been effective?"
This paragraph documents instances observed by the inspectors which indicate a lack of management effer.tiveness that appears to have resulted~in a failure to operate the facility consistent with the As Low As Is Reasonably Achievable (ALARA) criteria during 1985.
- 7. Exit Interview The inspector met with the licensee representatives denoted in Paragraph 1 at the conclusion of each site visit. The scope and findings of the inspection were summarized. The licensee representatives were informed of the apparent violations of NRC requirements discussed in this report.
The licensee indicated that the matters would be evaluated and appropriate actions would be taken as indicated in this report.
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