IR 05000334/1981021
| ML20041F006 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 02/08/1982 |
| From: | Baer R, Knapp P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20041F000 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 50-334-81-21, NUDOCS 8203160081 | |
| Download: ML20041F006 (20) | |
Text
.
.
.
U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION I
Report No. 50-334/81-21 Docket No. 50-334 License No. DPR-66 Priority
-
Category C
Licensee: Duquesne Light Company 435 Sixth Avenue Pittsburgh, Pe-', Ivania Facility Name: Beaver Valley Power Station, Unit 1 Inspection at: Shippingport, Pennsylvania Inspection conduct Augu_st 10-14, and 26-28, 1981
-
Inspector:
[
[M W
c
, A E-Baer, adiation Specialist date signed (
N.
C~
d te signed Approved by:
b h,_ _
C4 rO h/ k P. J. 'Knapp, Chief, Facility Radiological
/dafesigned Protection Section, Technical Inspection Branch Inspection Summary:
Inspection on August 10-14 and 26-28, 1981 (Report No. 50-334/81-21)
Areas Inspected:
Routine, unannounced inspection by a regional based inspector on transportation activities and radioactive waste management programs including:
management controls; selection of packages; preparation of packages for shipment; delivery of completed packages to carrier; receipt of packages; incident reporting; indoctrination and training program; audit program; record-keeping; NUREG-0578 Category A items.
The inspection involved 48 inspector hours on-site by one regional based inspector.
Results: Of the ten areas inspected no items of noncompliance were identified, r203160081 820225 PDR ADOCK 05000334
.
.
.
DETAILS 1.
Persons Contacted Mr. H.P. Williams, Station Superintendent
Mr. S.
Hall, Quality Control Engineer Mr. C.
Haney, Operations and Maintenance Instructor Mr. H.
Jenkins, Operations Foreman
Mr. J. A. Kosmal,
Radcon Supervisor
Mr. W.S. Lacey, Chief Engineer Mr. V.J. Linnenbom, Radiochemist Mr. F.J. Lipchick, Senior Compliance Engineer
Mr. A.J. Mizia, Senior Quality Assurance Engineer
Mr. J.D. Sieber Manager, Nuclear Safety and Licensing
Mr. S.
Sovick, Compliance Engineer Mr. J.
Vasello, Training Supervisor Mr. J.W. Wenkhous, Reactor Control Chemist
Other Personnel Mr. D.
Beckman, Sento,r Resident Reactor Inspector, USNRC Mr. J.
Hegner, Resident Reactor Inspector, USNRC The inspector also interviewed several other licensee and contractor employees including health physics technicians and operations personnel.
Denotes those persons present at the exit interview on August 14,
1981.
Denotes those persons present at the exit interview on August 14, and
August 28, 1981.
Denotes those persons present at the exit interview on August 28,
1981.
2.
Management Controls
.
The licensee has documented the management control for radioactive material management in the Beaver Valley Power Station (BVPS) Unit 1, Station Adminis-trative Procedures.
Chapter 2,Section IV F, specifies the Radiation Control Supervisor as the responsible individual for the radiological aspects of nuclear shipments leaving the site. Additional details of this responsibility are defined in Chapter 6 and in the Radiological Control Manual, Appendix 8.
The station operations group is responsible for the performance of operations
.
.
.
by operators and/or maintenance personnel relating to the processing, packaging, storage and shipment of radioactive solid waste. One individual, an operations supervisor, is assigned on a full time basis to radioactive waste operations. Other personnel are assigned as deemed necessary. At present this requires one operator approximately two days per week and one maintenance laborer five days per week.
The licensee has developed and implemented procedures for the various processes and details of the radioactive material handling program. Operating procedures incorporated in the BVPS Unit 1, Operating Manual, Chapter 18, Solid Waste Disposal System are:
18-E Solidification of Spent Resin 18-J Solidification of Liquid Wastes 18-N Liner Decontamination 18-Q Solid Waste Baler Operation 18-X Storage and Handling of Solid Waste Drums and Liners Radiological procedures incorporated in the BVPS Unit.I, Radiological Control Manual are:
RP-2.5 - Issue 1, Revision 3 Drumming of Solid Waste RP-3.2 - Issue 2 Labeling and Packaging Radioactive and/or Contaminated Material RP-3.3 - Issue 2 Receiving Radioactive Material RP-3.7 - Issue 1, Revision 2 Transfer of Material from a Contaminated Area or System RP-3.8 - Issue 1 Handling of Solid Radioactive Wastes RP-3.9 - Issue 2 Monitoring Vehicles
.
RP-3.10 - Issue 1, Revision 1 Estimating Curie Content of Radioactive Material RP-3.ll - Issue 1 Shipping Solid Radioactive Material for Burial - Drums RP-3.12 - Issue 1, Revision 2 Shipping Solid Radioactive Waste for Disposal - Liners RP-3.13 - Issue 1, Revision 1 Requirement Checklist for Shipping Greater than Type A Quantities of Radioactive Material RP-3.14 - Issue 1 EstimatingCurieContentofRadioactive
Material 65 ft Liners
., -
-
.
.
RP-3.15 - Issue 1 Estimating Curie Content of Filter Cartridges of Other Solid Objects
Immobilized in Cement Within 65 ft Liners RP-3.16 - Issue 1 Radioactive Shipment Seal Accountability RP-12.3 - Issue 1, Revision 2 Radioactive Shipment Records A documented program for indoctrination and training of personnel performing radioactive material handling is provided for by the following training programs:
-
System Qualification Standard, Chapter 18, Solid Waste Disposal System
.
Radiation Protection Module 6, Radiation Detection and Radioactive
-
Shipment The Operational Quality Assurance Department is responsible for planned and periodic audits of the radioactive waste management program.
The licensee has developed, as part of their Operational Quality Assurance Program Manual, the following procedures to provide guidance in implementing these audits:
,
OP-1 - Revision 4 Operational Quality Assurance Program OP-2 - Revision 5 Organization and Responsibilities OP-14 - Revision 3 Indoctrination and Training No items of noncompliance were identified.
3.
Selection of Packages The licensee's program for selection of packages was examined against the requirements of 10 CFR 71.12 and 71.54 and within the framework of the 00T requirements of 49 CFR Part 173.
The inspector reviewed the licensee's records for the procurement and reuse of packages used to transport radioactive material.
The licensee had not procured or fabricated any packages that required NRC Certification. Radio-active material shipments requiring the use of an NRC Certified package (outer container) were made in vendor-supplied containers. No containers were available during the inspection to determine conformance with the approval criteria, including 00T specification or Certificate marking requirements.
The inspector verified that the licensee had received prior approval from the NRC's Office of Nuclear Material Safety and Safeguards for all NRC Certified packages. The licensee had available current copies of the
-
_.
.
..
-
-
.
.
.
.
,
Certificate of Compliance, including the supporting safety analysis and cask handling procedures, for all type A and B cask he was authorized to use.
All shipments of radioactive material were made as full load shipments in transport vehicles consigned for exclusive (sole) use only.
No items of noncompliance were identified.
4.
Preparation of Packages for Shipment The licensee's program for preparation of radioactive material for shipment was reviewed against the requirements of 10 CFR 71.12, 71.31, 71.35, 71.53,-
71.54 and the following generally accepted codes and guides:
-
Regulatory Guide 7.1 - Administrative Guide for Packaging and Trans-porting Radioactive Material Regulatory Guide 7.4 - Leakage Tests on Package; for Shipment of
-
Radioactive Materials
-
ANSI N14.10.1 - Administrative Guide for Packaging and Transporting Radioactive Materials
-
ANSI N14.5 - Leakage Tests on Packages for Shipment of Radioactive Materials
-
45 CFR Parts 172 and 173 - Transportation The licensee has developed and implemented procedures for preparation of radioactive materials for shipment. These procedures (see list in paragraph 2) include requirements for a visual inspection of the package prior to filling or loading; how to close and seal the package; marking of the package weight and contents; labeling requirements appropriate for the type of package; and radiation and contamination limits for packages.
The licensee was not packaging radioactive material during this inspection.
The inspector noted by observation of the radioactive waste storage facility and procedure review that the licensee, for shipments of low specific
'
activity radioactive vaste, uses steel drums manufactured in accordance with DOT specification 17H (49 CFR 178.118).
Evaporator bottoms and spent demineralizer resins were solidified with concrete in steel containers.
No items of noncompliance were identified.
5.
Delivery of Completed packages to Carrier The licensee's program for delivery of completed packages to a carrier for transport was reviewed against the requirements of 10 CFR 71.55 and 49 CFR
'
-.-
-
-_ -
.
.
.
Parts 100 to 199. Activities for delivery of completed packages to a carrier were governed by previously mentioned procedures RP-3.11, RP-3.12, RP-3.13 and RP-12.3.
The inspector examined these procedures for consistency with regulatory raquirements and to determine whether they cnvered all aspects of the work being performed. Records to verify adherence to procedural requirements were also reviewed.
No items of noncompliance were identified.
6.
Receipt of Packages The licensee's program for the receipt of packages containing radioactive -
material was examined against the requirements of 10 CFR 20.205 and con-formance to procedure RP-3.3.
The inspector reviewed this procedure for compatibility with regulatory requirements and to determine if it covered all aspects of the work bhing carried out.
,
No items of noncompliance were identified.
7.
Incident Reporting The licensee uses contract carriers for the transportation of all radioactive waste packages. At each loading, the carrier was supplied with a written list of personnel and agencies to contact for any incident which may' occur during the course of transporting radioactive materials. Under instructions supplied with the list, the contract carrier's dispatcher must notify the licensee of all incidents and unanticipated delays encountered during transport activity.
Packages offered for transport were inspected prior to loading on the transport vehicle and during unloading.
There have been no instances'where
~
the effectiveness of the packages was substantially reduced during use.
~~
No items of noncompliance were identified.
8.
Indoctrination and Training Program The licensee's indoctrination and training program, as it pertained to the packaging of low level radioactive wJste for transport and burial, was examined against the provisions of IE Bulletin No. 79-19 and the licensee's response to this Bulletin.
The licensee implemented a training program and completed the initial training by January 1980.
Specific individuals received training in selected portions, according to their responsibilities, in the overall radioactive waste program.
e~,
,
'l
.I
.
.
,
_
-.
,
,
,
',~
- -
,
- .
,. -
"
,
,,
'
C
!'l
'
'
"'
" i
,
Operations personr.el rece6e a,nnual retrainiii<;,as part of Systems Qualifi-cation Standard, Chapter 18, Scl,id Waste Dispo' sal System.
The licensee h(d' sche iled a vender to present a training course on the
' preparation for transpbrt,and burial of radioactive waste. This course was
'
to be presented to radfat' ion control, quality assurance, quality control 7 and selected maintenance' personnel.
/.
No items of noncompliance were identified.
-
9.
Audit Program
'S
,
The licensee's audit function for the low-level radioactive waste transfer, packaging and transport activities was examined against the requirements of g 10 CFR 71 and IE Bulletin No. 79-19 and within the framework of the following generally accepted guidance:
,
,
'
Regulatory Guide 1.33 - Quality Ass,ura' ace Program Requirements
-
.
-
,
ANSI N18.7-1976 - Administrative, Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants
,
,
The inspector reviewed the' latest a6dit BV-1-81-07, dated March 9-12, 1981 of transportation activities conducted by the licensee. The followin,g areas were reviewed during the Audit and found to be satisfactory:
Labeling,storageandmonitorinh'ofradioactivewastepackages; shipment vehicle monitoring and inspection; waste packaging seals and closures; shipment records and worksheets DOT requirements and specifications for radioactive waste shipments The audit was conducte'd in accordance wit,h the licensee's written procedures listed in paragraph 2 of this report.,b deficiencies were identified during this audit.
I e
No items of noccompliance were identified.
av
.
10.
Recordkeeping f,
Theinspectorreviewedtherecord$ fall 29radioactivematerialshipments made by the licensee during the perfod June 25, 1979 through December 31, 1979, 188 shipments in 1980, and 40 shipments made through August 14, 1981, for compliance with the requirements of 10 CFR 71.62. The licensee maintains a record of all radioactive material shipments in excess of the tw6 ye'ar requirement.
The records were complete with all shipping document'ation',
including the shipping radiation survey data.
<<
'
!
!
No items of noncompliance were identified.
!
'
i
'
&
b
l /
i o,
)r y
< -,
-
._
_
.
_ _ - _
-
_
.
_
_
. _ _ _ _ _ _
.
.
11. TMI Lessons Learned - Category A Items i
]
{
As a result of the accident at Three Mile Island (TMI) and the significant safety concerns identified subsequent to the accident, the NRC's Office of Nuclear Reactor Regulation established a Lessons Learned Task Force to identify and evaluate safety concerns brought to light by the TMI accident.
The findings and recommendations of the Lessons Learned Task Force were published as NUREG-0578, in July 1979.
The recommendations were divided into ' Category A' items recommended to be implemented by January 1, 1980 and ' Category B' items recommended to be implemented by January 1, 1981.
NUREG-0578 made, among other recommendations, the recommendation that post-accident sampling be improved (Section 2.1.8.a), radiation monitor ranges be increased (Section 2.1.8.b), and in plant iodine instrumentation be improved (2.1.8.c).
Item 2.1.8.a. post-accident sampling capability, dealt with the necessity to take, handle and analyze, highly radioactive samples of the reactor
-
coolant and the containment atmosphere while, at the same time, maintaining personnel exposure as low as reasonably achievable and below specified maximum values. Chemical and radiological analyses were specified and the time within which collection and analyses were to be completed was set forth.
Item 2.1.8.b, increased range of radiation monitors, provided licensee's with minimum acceptable ranges of instruments and stated that capability for effluent monitoring of radioactivity was to be accomplished through the use of noble' gas effluent monitors and iodine sampling with charcoai cr other media. Two independent in-containment radiation level monitors were also called for.
Item 2.1.8.c, improved in plant iodine instrumentation, required that licensee's provide equipment, training, and procedures for accurately determining the airborne iodine concentration in-areas within the facility where personnel may be present during an accident.
Subsequent to the issuance of NUREG-0578, the NRC provided, in letters dated September 13, 1979, and October 30, 1979, additional clarification and requirements. On January 2, 1980, a Confirmatory Order was issued to appropriate licensee's to confirm their commitment to implement the applicable NUREG-0578 items (' Category A').
.
Additional NRC Requirements and Licensee Commitments i
NRC Letter of September 13, 1979 In a letter to all operating Nuclear Power plants dated September 13, 1979 from D. Eisenhut, NRR stated, "... we have concluded that all operating
,
reactor licensees should begin to implement the actions contained in NUREG-
..
.
.,
-
.,-..=
--
-
.
. -
.
.
0578... as soon as possible. Accordingly, please submit within 30 days of receipt of this letter, your commitment to meet these requirements on the implementation schedule contained in Enclosure 6."
The relevant portion of Enclosure 6 reads as follows:
Section Abbreviated Position Implementation No.
Title Description Category
,
2.1.8.a Post Accident Design review A
Sampling complete Preparation of A
revised procedures Implement plant B
modifications Description of pro-A posed modification 2.1.8.b High Range Radiation Installation B
Monitors complete 2.1.8.c Improved Iodine Complete A
Instrumentation Implementation Category A:
Implementation complete by January 1, 1980, or prior to OL, (operating license) which ever is later.
Category B:
Implementation complete by January 1, 1981.
The September 13 letter further states;
"Other Review Areas Enclosure 7 outlines the requirements developed to date resulting from the staff's Emergency Preparedness Studies.
Enclosure 8 provides the implementa-tion schedules for the emergency preparedness requirements which, you wi7 i note, includes three of the Lessons Learned Topics. We also require tha, you provide commitments to comply with each of the requirements of Enclosuie 7 in accordance with the implementation schedules shown in Enclosure 8.
Such commitments should be included in your letter due in 30 days of receipt of this letter."
The pertinent Section of Enclosure 7 states:
"Our near term requirements in this effort are as follows:
-.,_
. - _
.-
.
_ _--
_ ___-.
--. - - - - -
-
i
3 (2) Assure the implementation of the related recommendations of the Lessons
'
Learned Task Force involving instrumentation to follow the course of an accident and relate the information provided by this instrumentation to the emergency plan action levels. This will include instrumentation for post-accident sampling, high range radioactivity monitors, and improved in plant radioiodine instrumentation. The implementation of
the Lessons Learned Task Force's recommendations on instrumentation for detection of inadequate core cooling will also be factored into
,
the emergency plan action level criteria."
,
The pertinent Section of Enclosure 8 reads as follows:
l
" Enclosure '}
NEAR TERM EMERGENCY PREPAREDNESS IMPROVEMENTS IMPLEMENTATION SCHEDULE
,
Item Implementation Category 1/
.
1.
Implement certain short-term actions recommended
'
by Lessons Learned Task Force and use these in i
action level criteria.
2.1.8(a) Post-accident sampling Design review complete A
Preparation of revised procedures A
Implement plant modifications B
Description of proposed modification A
2.1.8(b) High range radioactivity monitors Methods for estimating release A
'
High range monitors B
2.1.8(c)
Improved in plant iodir.e instrumentation A-1/Category A:
Implementation prior to OL or by January 1, 1980
(see NUREG-0578)
,
i Category A : Implementation prior to OL or by mid-1980.
Category B:
Implementation by January 1,1981."
NOTE:
NUREG-0737, Clarification of TMI Action Plan Requirements, issued November 1980, changed the implementation date for 2.1.8(b) High Range Monitors (noble gas, iodine / particulate and containment high range) to January 1, 1982.
.-
_,
. _ _... _ _, _ _. _ _
_ _.
.
_
_-
_
- _
- _
_ __ _
_
. - _.. __
_
.
.
.
i J
,
Licensee Letter of October 22, 1979 f
The licensee replied to the September 13, 1979 letter with a letter dated October 22, 1979, one portion of which stated "The unit will be shut down for refueling in December and will not be ready to return to power until late next spring due to extensive planned modifications and also due to corpleting the requirements of numerous IB Bulletins, including 79-02 and 79-14."
In addition, the referenced letter requested, "Since the unit will not be in operation on January 1, 1980, we would like to obtain your approval to defer full compliance with the Implementation Category A items until the unit is ready to resume power operation."
The appropriate statements in the licensee's October 22, 1979 letter are as follows:
"2.1.8(a) Improved Post Accident Sampling Capability We are performing necessary design reviews and procedure preparation to comply with the Implementation Category A items subsequent to the refueling.
The Category B item requirements will be identified, engineered and scheduled upon the completion of the ongoing
,
review.
"
2.1.8(b)
Increased Range of Radiation Monitors We are conducting an engineering study to identify the equipment necessary to comply with this requirement. The procurement and
'
installation schedule will be established upon the completion of this study.
2.1.8(c)
Improved Plant Iodine Instrumentation We have ordered silver zeolite cartridges for our plant air sampling equipment at this time. We will be in compliance with this requirement as soon as these cartridges are received and installed."
NRC Letter of October 30, 1979 (Item 2.1.8.a)
In a letter to all operating nuclear power plants dated October 30, 1979 from Harold R. Denton, further clarification of NRC staff requirements was
-
provided with regard to 2.1.8.a Category A items. The attachment to this letter more precisely defined the analysis capabilities required for both primary coolant and containment atmosphere samples. The attachment stated, in part:
"In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions.
Procedures shall be
-
.
.
--
,
,. _
. -.
..
_
-.
._..
_
.
.
.
4 provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source term).
Both analyses shall be capable of being completed promptly; i.e., the boron sample analysis within an hour and the chloride sample analysis within a shift... Plant procedures for the handling and analysis of samples, minor plant modifications for taking samples and a design review and procedural modifications (if necessary) shall be completed by January 1, 1980."
The attachment listed matters which were to be considered in the design review.
It also specified the provisions which should be included in the
,
licensee's radiological sample :nalysis capability. These included provisions to:
1.
Identify and quantify isotopes of previously listed nuclide categories to a specified sensitivity 2.
Dilute samples where necessary to provide capability for measurement and reduction of personnel exposure 3.
Restrict background to provide a specified error value 4.
Maintain plant procedures which identify the analysis required, measurement techniques and provisions for reducing background The attachment further stated:
"In performing the review of sampling and analysis capability, consi-deration shall be given to personnel occupational exposure.
Procedural changes and/or plant modifications must assure that it shall be possible to obtain and analyze a sample while incurring a radiation dose to any individual that is as low as reasonably achievable and not in excess of GDC 19."
Licensee Letter on November 30, 1979 (Item 2.1.8.a)
l The licensee replied to the October 30 letter in a letter dated November 30, 1979. With regard to item 2.1.8.a. the licensee's letter stated:
"We are presently performing a comprehensive review of plant post accident sampling and analytical capabilities including the capability to perform chemical analysis of highly radioactive samples.
Plant procedures for the handling and analysis of such samples and all possible plant modifications to facilitate the required capability will be completed prior to the restart of the plant in 1980. A schedule for the completion of all other modifications which have been determined to be necessary will be forwarded as soon as a schedule for these activities is developed. All requirements as described in the clarifi-cations on pages 27 through 30 of Enclosure 1 to your October 30, 1979, letter will be addressed."
-
- -.
-.
.
-.
.
-
.
.
Licensee Letter on June 26, 1980 (Item 2.1.8.a)
The licensee's June 26, 1980 letter presented details on the operation of the improved post-accident sampling system (PASS). With regard to current status the licensee's letter stated:
"For the interim, until the PASS becomes operational, capability exists for taking samples of the reactor coolant system and containment atmosphere. Non pressurized reactor coolant san 1ples can be obtained through the normal sample path on a continuous flow with purge directed to the VCT or drains tank. Additionally, a sample loop is provided to be able to divert sample flow to a PERKIN-ELMER flameless atomic absorption spectrophotometer. This instrument is equipped with an automated sampling device that can accurately draw microliter samples for boron analysis. The sampler can also obtain small unpressurized samples or perform sample dilution for isotopic analysis, pressurized samples are ion exchanged to remove unnecessary fission products and then a fixed volume is captured in a high pressure liquid chromatograph sampling valve.
The sample would be processed by a gas chromatograph equipped with a series of two component columns. The first coluren is designed to remove water, the second column separates the various components.
The effluent is used for noble gas isotopic analysir. This method of sampling would analyze for hydrogen, oxygen,
'
nitrogen and noble gas.
Containment atmosphere samples are obtained utilizing a sample bomb located in the Hydrogen recombiner analyzer utilizing the analyzer only.
The bomb is then removed to the lab where it can be analyzed for H2, 02 and activity.
It is expected that utilizing these interim methods, samples can,be obtained within a relatively short period of time with results obtainable one-half hour after the samples are obtained.
Laboratory facilities are in areas of low contamination in an accident condition and are accessible. An uncertainty exists as to the effect of potentially high background radiation on the counting equipment although the detectors are shielded.
For this reason, alternate counting capabilities are being provided in the onsite Technical Support Center.
Procedures are presently being developed and will be implemented prior to the scheduled plant startup in July to address the interim method for obtaining and analyzing samples and provide for radiological control when sampling."
NRC Letter of October 30, 1979 (Item 2.1.8.b)
With regard to item 2.1.8.b, the attachment to the October 30, 1979 letter provided clarification of the need for increased range of radiation monitors
-
. _.
.
.
j
,
'
and discussed radioiodine and particulate release monitoring. The attachment stated in part:
"1.
Radiological Noble Gas Effluent Monitors A.
January 1, 1980 Requirements Until final implementation in January 1,1980, all operating reactors must provide, by January 1, 1980, an interim method for quantifying high level releases which meets the requirements of Table 2.1.8.b.1...
Methods are to be developed to quantify release rates of up to 10,000 Ci/sec for noble gasses" The table which dealt with both noble gas and radioiodine releases is set forth below:
" TABLE 2.1.8.b.1 INTERIM PROCEDURES FOR QUANTIFYING HIGH LEVEL ACCIDENTAL RADI0 ACTIVITY RELEASES Licensees are to implement procedures for estimating noble gas and radioiodine release rates i' the existing effluent instrumentation goes off scale.
Examples of major elements of a highly radioactive effluent release special procedures (noble gas).
Pre-selected location to measure radiation from the exhaust air, 1.
-
e.g., exhaust duct or sample line.
Provide shielding to minimize background interference.
2.
-
3.
-
Use of an installed monitor (preferable) or dedicated portable monitor (acceptable) to measure the radiation.
4.
-
Pre-determined calculational method to convert the radiation level to radioactive effluent release rate."
The attachment further stated,
'
"For assessing radioiodine and particulate releases, special procedures
'
must be developed for the removal and analysis of the radioiodine/
particulate sampling media (i.e., charcoal canister / filter paper).
Existing sampling locations are expected to be adequate; however, special procedures for retrieval and analysis of the sampling media under accident conditions (e.g., high air and surface contamination and direct radiation levels) are needed...
. _ - -
. - -
_ - _ -.
.
. _ -
. _
.
-
.. -.
_.
..
.
.
2.
Radiofodine and Particulate Effluents A.
For January 1,1980 the licensee should provide the following:
1.
System / Method description including:
'a)
Instrumentation to be used for analysis of the sampling media with discussion on methods used to correct for potentially interfering background levels of radio-activity, b)
Monitoring / sampling location.
c)
Method to be used for retrieval and handling of sampling media to minimize occupational exposure.
d)
Method to be used for data analysis of individual radionuclides in the presence of high levels of radio-active noble gases, e)
If normal AC power is used for sample collection and analysis equipment, an alternate back-up power supply should be provided.
If DC power is used, the source should be capable of providing continuous read-out for 7 consecutive days.
2.
Procedures for conducting all aspects of the measurement analysis including:
a)
Minimizing occupational exposure b)
Calculational methods for determining release rates c)
Procedures for dissemination of information d)
Calibration and frequency and technique" By January 1, 1981 the licensee was required to provide high range noble gas effluent monitors with specified capabilities and characteristics for each release path. Also by January 1, 1981 the licensee was required to have the capability to continuously sample and provide onsite analysis of the sampling media for radioiodine and particulate effluents.
Furthermore, by January 1, 1981 the licensee was required to provide two radiation monitor systems in containment which were documented to meet the require-ments of Table 2.1.8.b.3 of the attachment. However, as previously noted, these implementation dates were changed to January 1, 1982 by NUREG-0737.
._,_.
.
-
-..
--
__
_ _ _ _ _ _ __
_ _ _ _ _ _ _ _.
.
.
Licensee Letter on November 30,1979 (Item 2.1.8.b)
The licensee's response in their November 30 letter, with regard to Item 2.1.8.b stated:
" Interim methods shall be developed to quantify high level radioactive releases which meet the requirements of Table 2.1.8.b.1 prior to the restart of the unit in mid 1980.
We are continuing with the preliminary engineering and discussions with equipment vendors necessary to place orders for the equipment required to comply with the staff position as clarified on pages 26 through 36 and Tables 2.1.8.b.2 and 2.1.8.b.3 of Enclosure 1 of your October 30, 1979, letter. We shall provide you with a schedule for the completion of the Category B requirements as soon as possible after the placement of purchase at which time the promised delivery dates of the equipment will be established."
Licensee Letter on June 26, 1980 (Item 2.1.8.b)
The licensee's June 26, 1980 letter, with regard to item 2.1.8.b stated:
2.1.8.b.1
"The three normal release points for noble gas effluents will be monitored over the range of 10-7 uCi/cc to 10-5 uCi/cc (relative Xe-133) by January 1, 1981. An interim system will provide monitoring over the range of 10-7 uCi/cc to 10-5 uCi/cc. This interim system will be operational prior to startup from present outage.
The normal effluent release points are as follows:
1.
Process Vent (on top of cooling tower)
2.
Elevated Vent (on top of Reactor Containment Building) and 3.
Auxiliary Building Vent The Waste Gas Decay Tanks and the Main Condenser Air Ejector are released via the Process Vent.
The Supplementary Leak Collection and Release System (which maintains the Main Steam Valve Area, the Safeguard Area, the Cable Vault Area, the Pipe Tunnel Area and the Fuel Building Area at negative pressure) is exhausted via the Elevated Vent.
The Containment Purge Exhaust is normally via either the Auxiliary Building Vent or the Elevated Vent. (It may also be via the process Vent).
The ventilation exhaust of the Auxiliary Building, which includes the Radwaste Area, is via the Auxiliary Building Vent. The Post Accident Containment Hydrogen Purge Exhaust is elevated to the Reactor Building Containment and any release is via the Reactor Containment Building (elevated) vent.
- _ - - _ _ - - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _..-_._ _
. - - -
._.
-
-
.
.
Each of these effluent release points are monitored for noble gases by the existing BVPS-1 Radiation Monitoring System.
In addition, the process monitor portion of this system provides continuous surveillance of noble gas activity in many of these areas upstream of the effluent monitors.
The range of existing effluent monitors is 2.9 x 10-5 uCi/cc to 2.9 x 10-2 uCi/cc (Xe-133) for Elevated Vent and 7 x 10-6 uC1/cc to 7 x 10-2 uCi/cc (Xe-133) for Process and Auxiliary Building Vents.
NRC Letter of October 30, 1979 (Item 2.1.8.c)
With regard to item 2.1.8.c, the attachment to the October 30, 1979 letter stated the reasons why portable instruments were necessary for monitoring airborne iodine levels. This discussion further emphasized the concept presented in NUREG-0578; that it was essential to have a reasonably accurate determination of the concentrations of iodine to which personnel dealing with the incident are exposed.
The attachment further stated:
" Iodine Filters and Measurement Techniques A.
The following are short-term recommendations and shall be imple-mented by the licensee by January 1,1980. The 'icensee shall have the capability to accurately detect the presence of iodine in the region of interest following an accident. This can be accomplished by using a portable or cart-mounted iodine sampler with attached single channel analyzer...
B.
By January 1, 1981:
The licensee shall have the capability to remove the sampling cartridge to a low background, low contamination area for further
analysis... The licensee shall have the capability to measure accurately the iodine concentrations present on these samples and effluent charcoal samples under accident conditions."
Licensee Letter of November 30, 1979 (Item 2.1.8.c)
The licensee's November 30 letter regarding item 2.1.8.c stated:
" Procedures shall be developed and implemented prior to the movement
'
of fuel during the upcoming refueling to accurately determine airborne iodine concentration in areas within the facility where plant personnel
may be present during an accident. This will be accomplished using portable iodine samplers equipped with silver zeolite cartridges.
,
Counting shall be performed with a properly calibrated single channel analyzer or with a laboratory type multi channel analyzer. We are
.
developing the capability to remove the sampling cartridge to a low background, low contamination area for further analysis in accordance with Clarification B to this staff position. We plan to have this completed by January 1, 1931."
_
_ - - -
.
_. -
-
_
.
.
,
J
,
Licensee Letter of June 26, 1980 (Item 2.1.8.c)
The licensee's June 26, 1980 letter, with regard to it m 2.1.8.c stated:
" Beaver Valley utilizes portable or cart-mounted air samplers with silver zeolite cartridges for collection of airborne radioiodine in-plant. Both AC and DC operated units are available.
Normally, the AC units are used, but the DC units, which utilize a 12 volt battery on the same cart, are available if AC power is not. Extra batteries and
,
l battery chargers are available to assure that batteries are fully charged when needed.
,
.
The zeollte cartridge is monitored with an appropriate instrument and detector to determine the concentration of radioiodine.
Tests performed at Bcaver Valley Power Station confirm that zeolite has excellent
!
discrimination in favor of radioiodine and against noble gases.
This monitoring is performed in the field (at the closest area with a suitable background). Normally, the detector used in the field is an Eberline Instrument Model HP-210, pancake type G-M counter. With this detector, an air sample collected over as little as two minutes with AC operated unit, the HP 210 could detect 1 MPC of radiciodine on the zeolite.
(The battery operating unit has a lower sampling rate so five minutes of sampling is required to obtain the same sensitivity).
Single channel analyzers with scintillation detectors (Eberline Model SAM-2 with RD-22 probe) are also av.iilable and calibrated for radioiodine assessment of the silver zeolite cartridnes.
If the preliminary checks with the HP-210 G-M detectors wam ant, the SAM-2 is used to refine the assessment. Laboratory analysis with a Ge Li system is another alternative method of assessment.
SAM-2 and Gamma Spectrometry are also provided in a mobile van facility which is presently being readied for use. Additional laboratory facilities will be provided in the Technical Support Center facility.
Procedures are being developed for the following:
1.
Post acciden', radioiodine sampling in plant.
.
2.
Calibration of radiation detectors for radioiodine monitoring
'
of zeolite cartridges.
3.
Radiological controls for personnel during field sampling and radiotodine assessment.
4.
Monitoring and evaluation of concentration of radiciodine in plant with available instruments.
5.
Procedures storing, maintaining and handling zeolite cartridges.
Appropriate personnel will be trained in the use of the above
,
procedures."
,
..
._
--
-
-
.- -
---
-
-
.
.
Findings During the Inspection of August 10-14 and 26-28, 1981 The licensee's Beaver Valley Power Station Unit 1 facility was shut down for a major outage between November 30, 1979 and November 21, 1980.
Containment Atmosphere Sampling (2.1.8.a)
Regarding containment atmosphere sampling, the licensee's letter of June
"
26, 1980 stated:
... samples are obtained utilizing a sample bomb located in the hydrogen recombiner analyzer...".
The method actually used employs a quick-disconnect syringe fitting (septym) located at the inlet to the "A" hydrogen analyzer. A gas tight syringe is used to obtain a 2 cubic centimeter (cc) sample from the septym fitting.
The syringe is transported to the chemistry laboratory in a lead shield for analyses.
Primary Coolant Sampling (2.1.8.a)
The licensee had installed a shielded sampling station using reach-rod control valves, quick disconnect sample fitting, and a rheodyne chromatograph sample valve with two, 2 cubic centimeter sample loops to obtain a primary coolant sample (RCS). A lead shield is available to transport the sample from the sample station to the chemistry laboratory where analyses is performed.
The licensee has provided a sample loop with a Perkin-Elmer Model 5000 atomic absorption spectrophotometer (AA) for boron analysis at the reactor coolant sampling station.
The AA is preprogrammed to analyze a blank, a standard, clean out steps and a RCS sample. A system error of + 150 parts per million has been determined.
Details for obtaining and analyzing pressurized and unpressurized RCS and containment atmosphere samples are contained in procedure 9.3 of the chemistry manual. This procedure was available prior to the November 21, 1980, reactor start-up (Inspection Report 50-334/80-27).
Noble Gas Effluent Monitors (2.1.8.b(1))
The licensea has installed an intermediate and high range off-line noble gas monitor at each of the three normal effluent release points. These monitors, powered by an emergency electrical bus and used to supplement the normal radiation monitors, are all located in the Primary Auxiliary Building and near a phone for communication with the Control Room.
Procedures were available for normal and emergency operating conditions.
Iodine and Particulate Monitoring (2.1.8.b(2))
The intermediate and high range noble gas monitors referred to above also have a particulate and radiciodine prefiltration and sample collection holder assembly.
This assembly is fitted with quick disconnects and contains a particulate filter and silver zeolite radioiodine cartridge. Procedures
_
.
...
-
..
d
.
.
-
are available for obtaining samples during normal operation, (Radiological Control Manual (RCM) Chapter 4, procedure 2.8; and for emergency conditions, Chapter 5, procedure 1.1.)
The filter paper and silver zeolite cartridge are transported, in a lead shield if necessary, to the chemistry laboratory for analyses.
In-Plant Iodine Monitoring (2.1.8.c)
The licensee has five portable particulate and iodine air sample carts,
.;
i which include a battery power supply and additional support supplies for
"
emergency conditions.
These units supplement the normal air sampling
,
equipment used routinely within the plant.
Silver zeolite cartridges are used for radioiodine collection. These cartridges are transported to the
'
chemistry laboratory for analyses on either a multichannel analyzer or a SAM-2/RD-22 Iodine-131 counting system.
In Chapter 3, procedure 7.4, of
,
!
the RCM, a method for the rapid evaluation of radioiodine using an E 140N
or equivalent with an HP-210 probe is also explained.
The inspector had no further questions on this topic.
12. Exit Interview The inspector met with licensee management representatives (denoted in Paragraph 1) on August 28, 1981. The inspector summarized the purpose and scope of the inspection and the findings.
,
.-
-.. -
.