NUREG-0473, Forwards Draft,Rev 3 to NUREG-0472, Std Radiological Effluent Controls for Pwrs, Containing Draft Guidance for Radiological Effluent Tech Specs/Odcm Commitments,Per Generic Ltr 89-01
| ML20246J573 | |
| Person / Time | |
|---|---|
| Issue date: | 08/28/1989 |
| From: | Liza Cunningham Office of Nuclear Reactor Regulation |
| To: | Bellamy R, Dan Collins, Greger L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| RTR-NUREG-0472, RTR-NUREG-0473, RTR-NUREG-472, RTR-NUREG-473 GL-89-01, GL-89-1, NUDOCS 8909050205 | |
| Download: ML20246J573 (255) | |
Text
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AUS 2 8 983
' MEMORANDUM FOR: Ronald R. Bellany, Chief, EPRPB, DRSS, Region I Douglas M. Collins, Chief, EPRPB, DRSS, Region II
< L. Robert Greger, Chief. EPRPB, DRSS, Region III
( Bleine Murray, Chief RPSB, DRSS, Region IV l Gregory P. Yuhas, Chief, EPRPB, DRSS, Region Y FROM: LeMoine J. Cunningham, Chief Radiation Protection Branch Division of Radiation Protection
- and Emergency Preparedness w< Office of Nuclear Reactor Regulation
SUBJECT:
UPGRADED DRAFT GUIDANCE FOR THE "RETS/0DCM" COMMITMENTS IN ACCORDANCE WITH GENERIC LETTER 89-01 My memorandum dated February 3,1989 forwarded a copy of NUREG-0472, Revision 3, Draft 8 as initial guidance for the CONTROLS that will occupy the first section of the ODCM upon transfer to the ODCM of the RETS in accordance with GL 89-01.
We have now completed assenbly of the Standard Radiological Effluent Controls (SREC) for both PWRs and BWRs, and the documents to be published as NUREGs are going through the concurrence chain. We fully expect that these two documents will be approved ano formally published as NUREGs, but it may be a few months before they are available. In the meantime, the enclosed copies of Draft 9 of this recent model guidance are made available informally for information in the same way the draft model RETS have been made available since 1978.
Original signed by LE%ne ). Cunningham LeMoine J. Cunningham, Chief l Radiation Protection Branch Division of Radiation Protection and Emergency Preparedness Office of Nuclear Reactor Regulation
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- OFFSITE DOSE CALCULATION MANUAL GUIDANCE: A A
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AAAA**AARR*****RARARRAARRARARAAAR***At**AAARRAARARRAARRAAAAAARRA This compilation of Standard Radiological Effluent Controls (SREC) contains all of the controls required by Generic Letter 89-01, to be incorporated into a licensee's Offsite Dose Calculation Manual (ODCM) at the time the procedural details of the current Radiological Effluent Technical Specifications (RETS) are transferred out of the licensee's Technical Specifi-cations (TS). It has been developed by recasting the RETS of the most current Westinghouse Standard Technical Specifications (W STS) from the "LC0" format into the " Controls" format of an ODCM entry. Since the RETS guidance for Babcock and Wilcox and Combustion Engineering plants is identical to that for Westinghouse plants, the W SREC are applicable to all Pressurized Water Reactors.
The following W SREC provide the latest version of staff guid-ance, and document current practice in the operating procedures required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a and Appendix I to 10 CFR Part 50.
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.; H 06 f TABLE OF CONTENTS-o h
. FOREWORD ....................................................... 'l
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1-0
'I DEFINITI0NS ...................................................
3/4- LIMITING CONDITIONS FOR OPERATION & SURVEILLANCE REQUIREMENTS .
3/4 0-0 3/4.0 Applicability.................................................. 3/4 0 ,3/4 3 3/4.3 Instrumentation ...............................................
3/4.11 Radioactive Effluents ......................................... 3/4 11-1 3/4.12 Radiological-Environmental Monitoring ......................... 3/4 12-1 3/4- BASES .........................................................
B 3/4 0-0 5-O
- 5' DESIGN FEA1URES ...............................................
.6-D' 6 ADMI NI STRAT IVE CONTROLS ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- i. APPENDIX A: Radiological Assessment Branch Technical Position, Revision 1,' November 1979 APPENDIX B: General Contents of the Offsite Dose Calculation Manual APPENDIX C: Generic Letter 89-01 c
y SREC Contents
.- ' DRB FOREWORD RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS Licensee Technical Specification (TS) amendment requests for incorporation of i
Radiological Effluent Technical Specifications (RETS) pursuant to 10 CFR 50.36a and Appendix I to 10 CFR Part 50 were approved in the mid-1980s for most operating reactors licensed before 1979 (ors). Plants licensed after 1979 (NT0Ls), included the RETS as part of their initial Technical Specifications.
By November 1987, the RETS were implemented by all licensees of operating power reactors. Detailed Safety Evaluation Reports (SERs) documented the accept-ability of the plant-specific RETS of the ors, while the acceptance of the RETS for the NTOLs followed the regular pattern of the Standard Technical Specifi-cations (STS). Thus, for all operating plants, the compliance of the licensee with 10 CFR 50.36a and Appendix I to 10 CFR Part 50 is a matter of record.
l Early draf t revisions of model RETS, distributed to licensees in mid-1978, contained equations for dose calculations, setpoint determinations and meteoro-logical dispersion factors, as well as the procedural details for complying with Appendix I to 10 CFR Part 50. In later revisions, including Revision 2 used as the bench mark for the NRC staff's acceptance of OR RETS, the equations were removed and incorporated into an Offsite Dose Calculation Manual (ODCM) prepared by the licensee and provided to NRC for review along with the proposed RETS.
Early guidance for preparation of the Radiological Effluent Technical Specifi-cations (RETS) and Offsite Dose Calculation Manual (ODCM) was published in NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978. Copies of model RETS, however, have been available only in draft form as NUREG-0472, Revision 2, " Radiological Effluent Technical Specifications for PWRs," February 1, 1980; NUREG-0473, Revision 2,
" Radiological Effluent Technical Specifications for BWRs," February 1,1980; and ucceeding draft revisions. Staff guidance for the Radiological Environmental Monitoring Program is contained in the Radiological Assessment Branch Technical Position (RAB-BTP), originally issued in March 1978 and upgraded by Revision 1 in November 1979 as a result of the accident at Three Mile Island. This Revision I to the RAB-BTP was forwarded to all operating reactor licensees in November 1979 and remains in effect at the present time.
Since this BTP was never incorporated into the Regulatory Guide System, a copy is reproduced in this document as Appendix A. Even though it has been used extensively in reviewing ODCMs, guidance for the contents of the ODCM is found only in an appendix to a paper presented at an Atomic Industrial Forum confer-ente in 1981, and has had only informal distribution since that time.
OFFSITE DOSE CALCULATION MANUAL The potential for augmentation of a licensee's ODCM through transfer of the procedural details of the RETS following the guidance of Generic Letter 89-01, provides an opportunity to assemble in one set of documents the staff guidance for the ODCM.
6 W SREC 1
The current overview guidance for development of the ODCM was prepared origi-nally in July 1978 and revised in February 1979 after discussions with commit-tees of the Atomic Industrial Forum. This guidance was made generally l available as " Appendix B - General Contents of the Offsite Dose Calculation l Manual (ODCM) (Revision 1, February 1979)" to the paper authored by C. A. Willis and F. J. Congel, " Status of NRC Radiological Effluent Technical Specification Activities" presented at the Atomic Industrial Forum Conference on NEPA and Nuclear Regulation, October 4-7, 1981, Washington, D.C. A copy of this guidance that continues in ef fect to date, is reproduced in this document as Appendix B.
During the discussions leading up to the implementation of the RETS by the ors, it became important to record in a "living" document certain interpretations and understandings reached in these discussions. The ODCM thus became a repository for such interpretations, as well as for other information requested by the staff in connection with its evaluation of licensee's commitments and performance under 10 CFR 50.36a and Appendix I to 10 CFR Part 50.
TECHNICAL SPECIFICATION IMPROVEMENT PROGRAM Recently, the NRC staff has examined the contents of the RETS in relation to the Commission's Interim Policy Statement on Technical Specification Improve-ments. The staff has determined that programmatic controls can be implemented in the Administrative Controls section of the Technical Specifications (TS) to satisfy existing regulatory requirements for RETS. At the same time, the procedural details of the current TS on radioactive effluents and radiological environmental monitoring can be relocated to the Offsite Dose Calculation Manual (0DCM).
To initiate the change, new programmatic controls for radioactive effluents and radiological environmental monitoring are incorporated in the TS to conform to the regulatory requirements of 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50. The procedural details included in licensees' present TS on radioactive effluents, environmental monitoring, and associated reporting requirements will be relocated to the ODCH. Licensees will handle future changes to these procedural details in the ODCM under the administrative controls for changes to the ODCM. Detailed guidance to effect the transfer of the RETS to the ODCM is given in Generic Letter 89-01, repro-duced in its entirety as Appendix C.
GUIDANCE FOR THE TRANSFER OF RETS TO ODCM Enclosure 1 of Generic Letter (GL) 89-01 of Appendix B provides detailed guidance for the preparation of a license amendment request to implement the transfer of RETS to ODCM. Page 1 of the enclosure states:
"The NP.C staff's intent in recommending --- the relocation of procedural details of the current RETS to the ODCM is to fulfill the goal of the Commission Policy Statement for Technical Specification Improvements. It is not the staff's intent to reduce the level of radiological effluent control. Rather, this amendment will provide programmatic controls for RETS consistent with regulatory requirements and allow relocation of the procedural details of current RETS to the ODCM."
W SREC 2
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. '- *f Page 2 of Enclosure 1 states:
...the procedural details covered in the licensee's currer,t RETS, con-sisting of the limiting conditions for operation, their applicability, remedial actions, surveillance requirements, and the Bases section of the TS for these requirements, are to be relocated to the ODCM --- in a manner that ensures that these details are incorporated in plant operating procedures. The NRC staff does not intend to repeat technical reviews of the relocated procedural details because their consistency with the applicable regulatory requirements is a matter of record from past NRC reviews of RETS."
DISCUSSION For the purpose of the transfer described in GL 89-01 of Appendix B, the RETS will consist of the specifications from the STS listed in Enclosure 2 of Appendix B of GL 89-01. Licensees with nonstandard TS should consider the analogous TS in their format.
It is suggested that the most straightforward method of transferring a licensee's commitments in the RETS t; the ODCM in accordance with GL 89-01 is to recast the RETS in the licensee's present TS from the " Limiting Condition for Operation (LCO)" format of the TS into the " Controls" format of the ODCM ,
entry. The accompanying package provides an example of this recasting into ,
Standard Radiological Effluent Controls (SREC) from the model RETS for Pressurized Water Reactors (PWRs). This recasting is in format only. The TS pages have been transferred to the ODCM without change except for the substi-tution of " Controls" for "LC0". Plants that have RETS that closely follow the STS format will be able to use the accompanying examples directly as guidance.
For plants with nonstandard RETS the transfer of TS commitments to the ODCM should be made similarly page by page, again with the substitution of
" Controls" for "LC0".
SUMMARY
As part of the license amendment request for TS improvement relative to the RETS, a licensee confirms that the guidance of Generic Letter 89-01 has been followed. This guidance includes the following:
"The procedural details covered in the licensee's current RETS, consisting of the limiting conditions for operation, their applicability, remedial actions, surveillance requirements, and the Bases section of the TS for these requirements, are to be relocated to the ODCM --- in a manner that ensures that these details are incorporated in plant operating procedures."
The Standard Radiological Effluent Controls (SREC) compiled in this report document current staff practice in the operating procedures required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a and Appendix I to 10 CFR Part 50.
Thus they contain all of the controls required by Generic Letter 89-01, to be incorporated into a licensee's ODCM at the time the procedural details of the ,
current RETS are transferred out of the licensee's TS. i i
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9 SECTION 1.0 DEFINITIONS s
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I 1.0 DEFINITIONS y.
L The defined terms of this section appear in capitalized type and are applicable throughout these Controls.
ACTION' 1.1= ACTION shall be that part-of a Control that prescribes remedial measures required under designated conditions.
ANALOG CHANNEL OPERATIONAL TEST 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injecticn of a simulated signal into the channel as close to the sensor as' practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the setpoints are within the required range and accuracy.
CHANNEL CALIBRATION
- 1. 5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be
. performed by any series of sequential, overlapping,' or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK 1.6 A CHANNEL CHECK shell be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter. ,
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gf ' ~1-. - 1 DEFINITIONS I..
5 DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of ~ I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in
.[ Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites" or Table E-7 of NRC Regulatory Guide 1.109, Revision 1, ,
October 1977).
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- DEFINITIONS y
C FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall- correspond to the intervals defined in Table 1.1.
MEMBER (5) 0F THE PUBLIC-1.1F, MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant. This category does not include employees of the licensee,_ its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.
This category does include persons who use portions of the site for recre-ational~, occupational, or other purposes not associated with the plant.
OFFSITE DOSE CALCULATION MANUAL 1.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semiannual' Radioactive Effluent Release Reports required by TS 6.9.1.3 and 6.9.1.4 ,
W SREC 1-3
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T c' DRAFT DEFINITIONS' OPERABLE'- OPERABILITY 1.18 A system, subsystem, train, component or device.shall be OPERABLE or
< have.0PERABillTY when it is capable of performing its specified function (s),
and when all necessary attendant instrumentation, controls, electrical power,
. cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perforn its function (s) are also capable of performing their related support function (s).
OPERATIONAL MODE'- MODE 1.19 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.2 of the [ plant name] TS.
PURGE - PURGING 1.23 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
6 ySREC 1-4
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DEFINITIONS p- ,
i RATED T.ERMAL-POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of MWt.
- REPORTABLE EVENT 1
1.27 A REPORTABLE EVENT shall be any of those conditions specified'in l Section 50.73 of-10 CFR Part 50.
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l l- SITE BOUNDARY l-1.30 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.
W SREC 1-5
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. .:t SOURCE CHECK 1.33 A SOURCE CHECK shall be the qualitative assessment of channel response when'the channel sensor is exposed.to a source of increased radioactivity.
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1.35 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant..
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- DEFINITIONS _
UNRESTRICTED AREA 1.38 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.
VENTILATION EXHAUST TREATMENT SYSTEM 1.39 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal ladsorbers and/or HEPA filters for the purpose of removing iodines or particu-lates from the gaseous exhaust stream prior to the release to the environment.
Such a system is not considered to have any effect on noble gas effluents.
Engineered Safety Features Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.
VENTING 1.40 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not pro-vided or required during VENTING. Vent, used in system names, does not imply a VENTING process.
WASTE GAS HOLDUP SYSTEM 1.41 A WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System offgases from the Reactor Coolant System and providing for delay or holdup ,
for the purpose of reducing the total radioactivity prior to release to the environment.
D W SREC 1-7 l
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. TABLE 1.1
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FREQUENCY NOTATION
' NOTATION. FREQUENCY S At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W At least once per 7 days.
M -At least once per 31 days.
Q At least once per 92 days.
SA At least once per 184 days..
R At least once per 18 months.
S/U Prior to each reactor startup.
N.A. Not applicable.
!' 'P Completed prior to each release.
ySREC 1-8
s DRAFT TABLE 1.2 l~
OPERATIONAL MODES REACTIVITY- % RATED AVERAGE COOLANT THERMAL POWER
- TEMPERATURE MODE CONDITION, K,ff-
- 1. POWER OPERATION > 0.99 > 5% 1 350'F.
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- 2. STARTUP > 0.99 5 5% ' > 350*F
' 3. HOT STANDBY < 0.99 ,
0 . > 350*F
- 4. HOT SHUT 00WN < 0.99 0 350'F > T
> > 200 F avg
- 5. COLD SHUTDOWN < 0.99 0 1 200*F
- 6. REFUELING ** 5 0.95 0 1 140*F
- Excluding decay heat.
- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
W SREC 1-9
O SECTIONS 3.0 AND 4.0 CONTROLS AND SURVEILLANCE REQUIREMENTS W SREC 3/4 0-0
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i, 3/4 CONTROLS AND SURVEILLANCE REQUIREMENTS 3/4.0 ' APPLICABILITY' CONTROLS 3.0.1 Compliance with the Controls contained in the succeeding controls is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Control, the associated ACTION require-ments shall be met.
3.0.2 Noncompliance with a control shall exist when the requirements of the Control and associated ACTION requirements are not met within the specified time intervals. If the Control is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required. .
3.0.3 When a Control is not met, except as provided in the associated ACTION requirements, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action shall be initiated to place the unit in a MODE in which the control'does not apply by placing it, as applicable, in:
- a. At least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- b. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
- c. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Control. Exceptions to these requirements are stated in the individual controls.
This control is not applicable in MODE 5 or 6.
l 3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Control are not met and the as.sociated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in accordance with ACTION requirements when conformance to them permits I continued operation of the facility for an unlimited period of time. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications.
1 W SREC 3/4 0-1 i
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' APPLICABILITY:
SURVEILLANCE REQUIREMENTS 4.0.1 ' Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Controls unless otherwise stated in an individual Surveillance Requirement.
4.0.2- Each Surveillance Requirement shall be performed within the specified time interval with:
- a. A maximum allowable extension not to exceed 25% of the surveillance interval, but
- b. The c*ombined time interval for any three consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance interval.
4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the OPERABILITY requirements for a Control. The time limits of the ACTION requirements are applicable at the time it is identified
'that a Surveillance Requirement has not been performed. The ACTION requirements may be delayed for up to'24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to permit the completion of the surveillance when the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Surveillance Requirements do not have to be performed on inoperable equipment.
4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Control has been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual controls.
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DWH l INSTRUMENTATION RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION CONTROLS 3.3.3.10 In accordance with [ plant name] TS 6.8.4.g.1), the radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Control 3.11.1.1 are not exceeded. The Alarm / Trip Setpoints of these channels
- shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DDSE CALCULATION MANUAL (ODCM).
APPLICABILITY: At all times.
ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above control, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report pursuant to Control 6.9.1.4 why this inoperability was not corrected in a timely manner.
- c. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
Report all deviations in the Semiannual Radioactive Effluent Release Report.
SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 4.3-8.
W SREC 3/4 3-72
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R T t a t o i t a l t o - i S in s t d in r o i t o do m s t N ni a a l ni e o s at l t e a a I om w r i om t C o r i r w r C Mr d e u Mr a p ee ue u d e e a n B e W t m nt Bt s a n yT R e yT n o ea a a R e t G e t e e C Gn en e G ic d n i c c n r nr M d vi i 'm i vi i o p
s u
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C o u n
t a S( T(
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t S
i D
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i u i u t dA dA n o a . . . a . . o . . l . .
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R .
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d o e G r u N t n I n
i R o t O c n T -
o I r
C N e
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- N e e 3 E r U n i
3 L n F L E
F o L E t B n d A D e e T I u s U e l a Q n f f
b I i L L E s i
E t n t V n w n I e o T u d i o
C T l w p A N
- f o t 0 E s f l I M r E B e D U e S A R d e 'r p
R T r t o S o s t i N c a a r I e w r T R d e /
a n m r
y R e t G a i d l v i m A i u a t q e f c i t i a L S o y i l d n a . . o R a b d
e r
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i
,e .
DflAFT TABLE 3.3-12 (Continued)
ACTION STATEMENTS ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channelr, OPERABLE requirement, effluent releases via this pathway may continue provided that prior to initiating a release:
- a. At least two independent samples are analyzed in accordance with Control 4.11.1.1.1, and
- b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving.
Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 36 - With the number of channels OPERABLE less than required by the Minimum Ci.annels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are analyzed for radioactivity at a lower limit of detection of no more than 10 7 microcurie /ml:
- a. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the specific activity of the secondary coolant is greater than 0.01 microcurie / gram DOSE EQUIVALENT I-131, or
- b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the specific activity of the secondary coolant is less than or equal to 0.01 microcurie / gram DOSE EQUIVALENT I-131.
ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, grab samples are collected and analyzed for radio-activity at a lower limit of detection of no more than 10 7 microcurie /ml. ,
ACTION 38 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump perfor-mance curves generated in place may be used to estimate flow.
ACTION 39 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the radioactivity level is determined at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases.
W SREC 3/4 3-75
141
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d a v d e i m h i i m l t u a c t u a n a q e s c q e o R i t i a i t T L S D o L S d N w i e E o d r M l . . . a . . i U F a b c R a b u q
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- TABLE 4.3-8 (Continued)
TABLE NOTATIONS (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic l isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:
- a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or
- b. Circuit failure, or
- c. Instrument indicates a downscale f ailure, or ]
i
- d. Instrument controls not set in operate mode.
(2) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- a. Instrument indicates measured levels above the Alarm Setpoint, or
- b. Circuit failure, or
- c. Instrument indicates a downscale failure, or i
- d. Instrument controls not set in operate mode.
(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
(4) CHANNEL CHECK shall consist of verifying indication of flow during periods i of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic, or batch releases are made. i l
l l
i W SREC 3/4 3-78 l
1
4
- O EH INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION CONTROLS
'3.3.3.11_ In accordance with [ plant name) TS 6.8.4.g.1), the radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Control 3.11.2.1 are not exceeded. The Alarm / Trip 5etpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.
APPLICABILITY: As shown in Table 3.3-13 ACTION:
- a. With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip 5etpoint less conservative than required by the above control, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative,
- b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13. Restore the inoperable instrumentation to OPERABLE status 'withir, 30 days and, if unsuccessful, explain in the next Semi-annual Radioactive Effluent Release Report pursuant to Control 6.9.1.4 why this inoperability was not corrected in a timely manner.
- c. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
Report all deviations in the Semiannual Radioactive Effluent Release Report.
SURVEILLANCE REQUIREMENTS 4.3.3.11 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the f requencies shown in Table 4.3-9. I W SREC 3/4 3-79
N
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l mt B S a es A U t u T O h sa E c yh S u S x A m s E G e r t r t o r s nr o r E s t o m ee t o V y i t e Vs i t I S n i t n n i T o n s ne o n C n M o y wd M o A o r M S on r M 0 i y e do y e I t t l r e t wC t l r e D a i p o t n o i p o t A l v m t a e l l v m t a R i i r a i R V Ba i r a i R t t e S n e t e S n n c l o w d rS c l o w e A p e M o n o A p e M o T V m t l a td m t l N 's a a e F t
an ra s
a S a a t
e F E a a S l t r s G l
u a r M e G u a el e
U r e c R e u nG e c R R A e n i l a e e n i l p
T l i t w p h Ge l i t w S e b d r o m x n b d r o m N t o o a l a E mi o o a l a I s N I P F S ab N I P F S a r er w e t u d h ST a . . . . . t . . . . .
R a b c d e O a b c d e
_ 8 9
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(( l l
..+ .
DRA7 TABLE 3.3-13 (Continued)
TABLE NOTATIONS
- At all times.
- 'Du' ring WASTE GAS HOLDUP SYSTEM operation.
ACTION STATEMENTS ACTION 45 - With the number of-channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment provided that prior to initiating _the release:
- a. At least two independent samples of the tank's contents are analyzed, and
- b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup.
Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 46 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
ACTION 47 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 48 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway.
ACTION 49 - NOT USED ACTION 50 - NOT USED i
W SREC 3/4 3-85
r .!
c?
I DRAFT 1 '
TAB'? 3.3-13 (Continued)
TABLE NOTATIONS (Continued)
' ACTION 51 - With the number lof channels,0PERABLE less'than required by the-Minimum Channels OPERABLE > requiremerit, effluent releases via the affected pathway may continue provided samples are contin-uously collected with auxiliary sampling' equipment as required in~iable 4.11-2.
L ' ACTION 52 - NOT USED r.
W SREC 3/4 3-86
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n M CC A.
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i t R UH N N N N A. A. A.
n T OC M N N N N o S S C N t
( I 9 G
- N 3 I L
. R EK 4 O NC D D T NE W W D D D W W E I AH
'L N HC B O C A M T
T N
E U n n r L o r o o r F i F t o i t o a t t i t E n l i a i n
S i n l o o U t o i M r O n r M t M E e e n y e S V l r e e t l r e A p o t V i p o t G g) m t a v m t a nd r a i R a i r a i R E i e e S n e t e S n V du l o w r c l o w I l n p e M o A A p e M o T ii m t l m t l C ut a a e F e s a a e F A Bn S l t g a S l t D o u a r a G u a r I yC e c R e r e c R e D r( n i l o e n i l p
A a i t w p t l i t w R i m d r o m Sm b d r o m T l e o a l a e o o a l a N it I P F S l t N I P F S E xs es M uy . . . . uy . . . .
U AS b c d e FS a b c d. e R
T S
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I 6 7
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M L E A
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U ON1 S 2 2 Q LDTE ( A. A. Q Q ( A. A.
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A L
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AB R N N R N N R R N HI -
O CL I A T C
) A d T e N u E EK . . . .
n M CC . . . .
U RE A. A. A. A.
i t R UH A. A. A. A.
n T OC M N N N N M N N N N o S S _
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- N 3 I L
. R EK O NC 4
T NE D W W D D D W W D D E I AH L N HC B O C A M T
T N m t E e s n U t s r m e r L
y o r e Vl o r F
F S t o t a t o E i t s ne i t n n i y wS n i S o o n S o o n U M o dd M o O
i t r M t n
wn oa y r
e M
E a y e r e S l t l r e e l l t l p o t A i i p o t V BGt i a
G t v m t a s v m t n i r a i R d reu i r a i R E e t e S n n ona t e S n V V c l o w a tih c l o w e M o p e M o abx A p I A t rrE m t l T a m t l s eu s a a e F C e s a a e F a
A r a S l t u nTr S l t r
A G u a r a e e G u a 0
e c R e h G ,s 'e c R e I
D e e n i l x: mn e n i l A t l i t w p E s mee l i t r
w o
p m
R s b d r o m a atd b d a
T a o o a l a r esn o o a l S
N w N I P F S eh t yo N I P F E d h c SSC M a . . . . t u . . . . .
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9 I 8 It$W wA w
~
j TABLE 4.3-9 (Continued)
TABLE NOTATIONS
- At all times.
(1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exists;
- a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or b.. Circuit failure, or
- c. Instrument indicates a downscale failure, or
- d. Instrument controls not set in operate mode.
(2) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of c,ie following conditions exists:
- a. Instrument indicates measured levels above the Alarm Setpoint, or
- b. Circuit failure, or
- c. Instrument indicates a downscale failure, or
- d. Instrument controls not set in operate mode.
(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
1 ~
4 i
ySREC 3/4 3-92 l
4
.- 'o j
[lilill:lr 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION l
CONTROLS 3.11.1.1 In accordance with [ plant name) TS 6.8.4.g.2) and 3), the concentra-I tion of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see. Figure 5.1-3) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concen-tration shall be limited to 2 x 10 4 microcurie /ml total activity.
APPLICABILITY: At all times.
ACTION:
- a. With the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS exceeding the above limits, immediately rr " - the concentration to within the above limits.
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.1.1 Radioactive liquid wastes shall be sampled and analyzed according to the sampiing and analysis program of Table 4.11-1.
4.11.1.1.2 The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of Control 3.11.1.1.
I W SREC 3/4 11-1
DRAFT TAPLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY (LLD)(1)
TYPE FREQUENCY FREQUENCY ANALYSIS (pCi/ml)
- 1. Batch Waste P P Release Each Batch Each Batch Principal Gamma 5x10 7 Tanks (2) Emitters (3) 1-131 1x10 6 a.
P M Dissolved and 1x10 5 One Batch /M Entrained Gases (Gamma Emitters) b.
P M H-3 1x10 5 Each Batch Composite (4)
Gross Alpha 1x10 7 c.
P Q Sr-89, Sr-90 5x10 8 Each Batch Composite (4)
Fe-55 1x10 6
- 2. Continuous W Principal Gamma 5x10 7 Releases (5) Continuous (6) Composite (6) Emitters (3)
I-131 1x10 6 a.
M M Dissolved and 1x10 5 Grab Sample Entrained Gases (Gamma Emitters) b.
" H-3 1x10 5 Continuous (6) Composite (6)
Gross Alpha 1x10 7 l c.
9 (6)
Sr-89, Sr-90 5x10J Continuous (6) Composite Fe-55 1x10 6 W SREC 3/4 11-2 )
~~
l' M ,
h , ... = _
y 7
. TABLE NOTAT70NS L I )The LLD is; defined, for purposes of these. controls,-as the smallest concentration ofiradioactive material in a sample.that will-yield a net.
~
,' : count, above system background, that will_be detected with 95% probability with:only 5% probability of falsely concluding that a blank observation:
represents a "real" signal.
For a' particular measurement system,' which may' include radiochemical-separation: .
8 b
LLD =
E V + 2.22 x'108 Y+ exp (-Aat) j Where:
4 LLD.= the "a priori" lower limit of detection (microcurie per unit mass or volume),.
sg = the standard deviation of-the background counting rate or of
-the. counting rate of a blank sample as appropriate (counts per minute),
E = the counting efficiency (counts per disintegration),
-V = the sample size (units of mass or volume),
2.22 x 108 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield,_ when applicable, A = the radioactive decay constant for the particular radionuclides (sec 1),:and at = the elapsed time between the midp,oint of sample collection and the tin'e of counting (sec). +
Typical values of E, V, Y, and at should be used in the calculation.
It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit-for a particular measurement.
.(2)A' batch release'is the discharge of liquid wastes of a discrete volume.
Prior to sampling for analyses, each batch shall be isolated, and then
' thoroughly mixed ,y a method described in the ODCM to assure representative sampling.
L 1 .
.W SREC 3/4 11-3
. E _m.____ .__.m. _ _ _ _ _ . . ._, _ _ . _ _ , _ . _ . _ , _ _ . _ , ,
DRAFT 1 i
TABLE 4.11-1 (Continued) ,
TABLE NOTATIONS (Continued) f 1
(3)The principal gamma emmiters for which the LLD control applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, and Ce-141. Ce-144 shall also be measured, but with an LLD of 5 x 10 6 This list does not mean that only these nuclides are to !
be considered. Other gamma peaks that are identifiable, together with those of the above. nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Control 6.9.1.4 in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974.
l (4)A composite sample it one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.
(5)A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release.
(0)To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.
I I I 0
i W SREC 3/4 11-4 1
- J I
t- J
i DMFI RADIOACTIVE EFFLUENTS DOSE CONTROLS 3-3.11.1.2 In accordance with [ plant name) TS 6.8.4.g.4) and 6.8.4.A.5), the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Figure 5.1-3) shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrems to the whole body and to less than or equal to 5 mrems to any organ, and l
- b. During any calendar year to less than or equal to 3 mrems to the whole body and to less than or equal to 10 mrems to any organ.
APPLICABILITY: At all times.
ACTION:
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Control 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been 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. This Special Report shall also include: (1) the results of radiological analyses of the drinking water source, and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR Part 141, Safe Drinking Water Act.*
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.2 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance
~
with the methodology and parameters in the ODCM at least once per 31 days. ,
I i
- The requirements of ACTION a.(1) and (2) are applicable only if drinking water supply is taken from the receiving water body within 3 miles of the plant discharge. In the case of river-sited plants this is 3 miles downstream only.
l W SREC 3/4 11-5
^
DMFT RADIOACTIVE EFFLUENTS L1 QUID RADWASTE TREATMENT SYSTEM CONTROLS 3.11.1.3 In accordance with [ plant name] TS 6.8.4.g.6), the Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity when the projected doses due to the liquid effluent, from each unit, to UNRESTRICTED AREAS (see Figure 5.1-3) would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31-day period.
APPLICABILITY: At all times.
ACTION:
- a. With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the Liquid Radwaste ,
Treatment System not in operation, prepare and submit to the Commis-sion within 30 days, pursuant to Control 6.9.2, a Special Report that includes the following information: .
- 1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.3.1- Doses due to liquid releases from each unit to UNRESTRICTED AREAS __
shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Liquid Radwaste Treatment Systems are not being fully utilized.
4.11.1.3.2 The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting Controls 3.11.1.1 and 3.11.1.2.
W SREC 3/4 11-6
__-__-_--.-------.---.----7.---
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W SREC 3/4 11-7 f
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' RADIOACTIVE EFFLUENTS-l 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE CONTROLS 3.11.2.1 In accordance with [ plant name) TS 6.8.4.g.3) and 7), the dose rate due to radioactive materials released in gaseous effluents from the site to areas at and_ beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following:
- a. For noble gases: Less than or equal to 500 mrems/yr to the whole body and less than or aqual to 3000 mrems/yr to the skin, and
- b. For Iodine-131, for Iodine-133, for tritium, and for all radio-nuclides in particulate form with half-lives greater than 8 days:
Less than or equal to 1500 mrems/yr to any organ.
APPLICABILITY: At all times.
ACTION:
- a. With the dose rate (s) exceeding the above limits, immediately restore the release rate to within the above limit (s).
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM.
4.11.2.1.2 The dose rate due to Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining l l representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2.
l l
t -ySREC 3/4 11-8 L
l _ _. _ ________ _______-__-_-- _ - _ -
)
I
(
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T(l 1 t I m 2 1 MN/ 4 4 8 4 4 8 4 1 1 1 I Oi - - - - 8 - - - - t.
LI C 0 0 0 0 - 0 0 0 0 0 0 0 T p 1 1 1 1 0 1 1 1 1 1 1 1 RC( x x x x 1 x x x x x x x EE - 1 1 1 1 x 1 1 1 1 1 1 1 WT 1 OE LD
) ) ) ) ) )
M 2 2 2 2 2 2 A ( ( ( ( ( (
R s s s s s s G r r r r r r O e e e e e e R t t t t t t P t t t t t t i i i i i i S m m m m m m I S E E E E E E S I Y S a a a a a a L Y m m m m m A L m m m a
m a
m a
m a
N FA a a 0 A ON G G ) G ) G ) G G a 9 A e e e h -
D E l l d l d l d l l p r N PY a a i a i a i a a l S A YT p p x p x p x p p A TI i i o i o i o i i ,
G V c c ( c ( c ( c 1 c s 9 2 N I n n n n n 3 n s 8
- I T i i 3 i 3 i 3 i 1 i o -
1 L C r r - r - r - r - r r r 1 P A P P H P H P H P 1 P G S
. M e 4 A l -
S ) rp rp E 3 am a L E I e P a P B T E t S A S Y k G ) ) ) a e e T A SC n R 3 7 l 7 l t e ,t
-W MIN USE PT a
PP U
M
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g M M
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yoe a ( u yce Msa Qsa it i S MYU cl il ol ol U ILQ h h rp t p pu pu O NAE c c am rm aa mc mc E I NR a a h a oi oi S MAF E E CS PS Ct Ct A
G ) ) ) )
)
E 3 6 6 6 6 V e ( e e e e ( ( (
s
(
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l El l l l s s T Y k p G p p p p u u u u C
A GC NN nm aa Rm U a
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a m
a m
a ', o u
o u
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0 IE TS PS I S S S n n n n I LU P P ' i i i i D PQ hb hb 3 b ) b b t t t t A ME ca ca 3 a 5 a a n n n n R AR ar ar ( r ( r r o o o o SF EG EG MG MG MG C C C C et E e e t n s P g g e s e e ,e Y a r g n aV p .v T r u t a o w y1 o o P n r i ydB T b E t e o t rag na S S t V t a aRS ei A n S. l i
,,r s s .
E s e t i l ad3 L a m n l at i gae ee E G nt a e e r. xd eh l td R in l ure ul rt esn ek ae P FAV ABAO Ri a S t n tv l U sa n l ,
O aT or . .
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EW C o a b Aa2 S
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- ~
TABLE 4.11-2 (Continued)
TABLE NOTATIONS (1)The LLD is defiied, for purposes of these controls, as the smallest conceritration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability.
with only 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system, which may include radiochemical separation:
4* S b
LLD =
E V 2.22 x IO S Y - exp (-Aat)
Where:
LLD = the "a priori" lower limit of detection (microcurie per unit mass or volume),
sb = the standard deviation of the background counting. rate or of the counting rate of a blank sample as appropriate (counts per minute),
E = the counting ef ficiency (counts per disintegration),
V = the sample size (units of mass or volume),
2.22 x 108 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclides (sec 1), and at = the elapsed time between the midpoint of sample collection and the time of counting (sec).
Typical values of E, V, Y, and at should be used in the calculation.
It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a_ posteriori (after the fact) limit for a particular measurement.
W SREC 3/4 11-10
5 r
.j 1, r is T h a[./P' l=
TABLE 4.11-2 (Continued) l-l' TABLE NOTATIONS (Continued) l'
- (2)The principal gamma emitters for which'the LLD control applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and ;
Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, '
I-131, Cs-134, Cs-137, Ce-141 and Ce-144 in Iodine and particulate
. releases. This list does not mean that only these nuclides are to'be considered. Other gamma peaks that are identifiable, together with those !
of the above'nuclides, shall also.be analyzed and reported in the l Semiannual Radioactive Effluent Release Report pursuant to Control. 6.9.1.4 in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June'1974.
(3) Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period.
(4) Tritium. grab samples shall be taken at least once per 2'4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> when the refueling canal is flooded.
(5) Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is ;
in the spent fuel pool.
(6)The ratio of the sample flow rate to the sampled stream flow rate shall be i known for the time period covered by each dose or dose rate calculation made in accordance with Controls 3.11.2.1, 3.11.2.2, and 3.11.2.3.
( ) Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing or after removal from sampler.
l Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least 7 days following each shutdown, startup, or THERMAL POWER change exceeding i 15% of RATED THERMAL POWER within a 1-hour period and analyses shall be !
completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLDs may be increased by a factor of 10.
This requirement does not apply if: (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the reactor coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that i effluent activity has not increased more than a factor of 3. l i
W SREC 3/4 11-11
DRAFT RADI0 ACTIVE EFFLUENTS DOSE - NOBLE GASES CONTROLS 3.11.2.2 In accordance with [ plant name) TS 6.8.4.g.5) and 8), the air dose due to nobin gases released in gaseous effluents, from each unit, to areas at and beyond Lhe SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and
- b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.
APPLICABILITY: At all times.
ACTION
- a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Control 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been 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.
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. __
y SREC 3/4 11-12
l l
RADI0 ACTIVE EFFLUENTS DOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADI0 ACTIVE MATERIAL IN PARTICULATE FORM CONTROLS 3.11.2.3 In accordance with [ plant name] TS 6.8.4.g.5) and 9), the dose to a MEMBER OF Tile PUBLIC from Iodine-131, Iodine-133, tritium, and all radio-nuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following:
- a. During any calendar quarter: Less tha, or equal to 7.5 mrems to any organ and,
- b. During any calendar year: Less than or equal to 15 mrems to any organ.
APPLICABILITY: At all times.
ACTION:
- a. With the calculated dose from the release of Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Control 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) rnd defines the corrective actions that have been 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.
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.3 Cumulative dose contributions for the current calendar quarter and current calendar year for Iodine-131, Iodine-133, tritium and radionuclides in particulate forn with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.
W SREC 3/4 11-13
DUT RADIOACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT SYSTEM CONTROLS 3.11.2.4 In accordance with [ plant name] TS 6.8.4.g.6), the VENTILATION EXHAUST TREATMENT SYSTEM and the WASTE GAS HOLDUP SYSTEM shall be OPERABLE and appropriate portions of these systems shall be used to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) would exceed:
- a. 0.2 mrad to air from gamma radiation, or
- b. 0.4 mrad to air from beta radiation, or
- c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.
APPLICABILITY: At all times.
ACTION:
- a. With radioactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Control 6.9.2, a Special Report that includes the following information:
- 1. Identification of any inoperable equipment or subsystems, and the reason for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.4.1 Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordarce with the methodology and parameters in the ODCM when Gaseous Radwaste Treatment Systems are not being fully utilized.
4.11.2.4.2 The installed VENTILATION EXHAUST TREATMENT SYSTEM and WASTE l GAS HOLDUP SYSTEM shall be considered OPERABLE by meeting Controls 3.11.2.1 and 3.11.2.2 or 5.11.2.3.
W SREC 3/4 11-14
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I W SREC 3/4 11-16
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RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE CONTROLS 3.11.4 In accordance with [ plant name] TS 6.8.4.g.11), the annual (calendar year) dose ur dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.
APPLICABILITY: At all times.
ACTION:
I
- a. With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Control 3.11.1.2a., 3.11.1.2b., 3.11.2.2a., 3.11.2.2b., 3.11.2.3a., or 3.11.2.3b., calculations shall be made including direct radiation contributions from the units (including outside storage tanks etc.) to determine whether the above limits of Control 3.11.4 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Control 6.9.2, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits.
This Special Report, as defined in 10 CFR 20.405(c), shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes
~
the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release condition result-
. ing in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accor-dance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. _
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Controls 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the methodology and parameters in the ODCM.
4.11.4.2 Cumulative dose contributions from direct radiation from the units (including outside storage tanks etc.) shall be determined in accordance with the methodology and parameters in the ODCM. This requirement is applicable only under conditions set forth in ACTION a. of Control 3.11.4.
! y SREC 3/4 11-18 ;
yy : ,;
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING.
DRUT L
3/4.12.1 MONITORING PROGRAM CONTROLS 3.12.1 In accordance with [ plant name) TS 6.8.4.h.1), the Radiological Environmental Monitoring Program shan De conducted as specified in .
Table 3.12-1.
APPLICABILITY: At all times.
ACTION:
- a. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Control 6.9.1.3, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
- b. With the level-of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days, pursuant to Control 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective 2ctions to be taken to reduce radioactive effluents so that the potential annual dose
- to a MEMBER OF THE PUBLIC is less than the calendar year limits of Controls- 3.11.1.2, 3.11.2.2, or 3.11.2.3. When more than
~
one of the radionuclides in Table 3.12-2 are detected in the sampling
. medium, this report shall be submitted if:
concentration (1) concentration (2) + ***> 1.0
. reporting level (1) , reporting level (2) -
When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose
- to a MEMBER OF THE PUBLIC from all radio-nuclides is equal to or greater than the calendar year limits of Control 3.11.1.2, 3.11.2.2, or 3.11.2.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report required by Control 6.9.1.3.
- The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.
i L
W SREC 3/4 32-1 ,
[ gyg RADIOLOGICAL ENVIRONMENTAL MONITORING CONTROLS ACTION (Continued)
- c. With milk or fresh leafy vegetation samples unavailable from one or more of the sample locations required by Table 3.12-1, identify specific locations for obtaining replacement samples and add them within ?0 days to the Radiological Environmental Monitoring Program given in the ODCM. The specific locations from which samples were unavailable may then be deleted from the monitoring program. Pursuant to Control 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table for the ODCM reflecting the new location (s) with supporting information identifying the cause of the unavailability of samples and justifying the selection of the new location (s) for obtaining samples.
- d. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected !
pursuant to Table 3.12-1 from the specific locations given in the table and figure (s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 3.12-1 and the detection capabilities required by Table 4.12-1.
9 J
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DMFT TABLE 3.12-1 (Continued)
TABLE NOTATIONS (1) Specific parameters of distance and direction sector from the centerline of one reactor, and additional description where pertinent, shall be pro-
.vided for each and every sample location in Table 3.12-1 in a table and figure (s) in the ODCM. Refer to NUREG-0133, " Preparation of Radiological Ef fluent Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Technical Position, Revision 1, November 1979. Deviations are permitted from the required sampling schedule ~ if specimens are unobtainable due to circumstances such as hazardous conditions, seasonal unavailability, and malfunction of auto-matic sampling equipment. If specimens are unobtainable due to sampling equipment malfunction, effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environ-mental Operating Report pursuant to Control 6.9.1.3. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pethway in question and appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program given in the ODCM. Pursuant to Control 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table for the ODCM reflect-ing the new location (s) with supporting information identifying the cause of the unavailability of samples for the pathway and justifying the selec-tion of the naw location (s) for obtaining samples.
(2) One or more instruments, such r.s a pressurized ion chamber, for measuring and recording dose rate contiroously may be used in place of, or in e.ddi-tion to, integrating docimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters.
Film badges shall not be used as dosimeters for measuring direct radiation.
(The 40 stations is not an absolute number. The number of direct radiation monitoring stations may be reduced according to geographical limitations; e.g., at an ocean site, some sectors will be over water so that the number of dosimeters may be reduced accordingly. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the speci-fic system used and should be selected to obtain optimum dose information with minimal fading.)
(3) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.
m W SREC 3/4 12-7
6 .- -
TABLE 3.12-1 (Co ginued)
TABLE NOTATIONS (Cor.tinued)
(4) Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.
(S) The " upstream sample" shall be taken at a distance beyond significant influence of the discharge. The " downstream". sample shall be taken in an area beyond but near the mixine zone. " Upstream" samples in an estuary must be taken far enough upstream to be beyond the plant influence. Salt water shall be sampled only when the receiving water is utilized for recreational activities.
(6) A composite sample is one in which the quantity (aliquot) of liquid sampled is proportional to the quantity of flowing liquid and in which the method of sampling employed results in a specimen that is representative of the liquid flow. In this program composite' sample aliquots shall be collected at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative-sample.
(7) Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharp'.
properties are suitable for contamination.
(8) The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM.
(4). If harvest occurs more than once a year, sainpling shall be performed during each discrete harvest. If harvest occurs contircously, sampling shall be monthly. Attention shall be paid to including samples of tuberous and root food products.
l W SREC 3/4 12-8
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.F TABLE 4.12-1 (Continued)
TABLE NOTATIONS (1)This list does'not mean that only these nuclides are to be considered.
Other peaks that are identifiable, together~with those of the above nuclides, shall also be analyzed and repor*# in the Annual Radiological Environmental Operating Report pursuant to control 6.9.1.3.
(2) Required detection capabilities for thermoluminescent dosimeters used-for environmental measurements shall be in accordance with the recommenda-tions of Regulatory Guide 4.13.
(3)The LLD is defined, for purposes of these controls, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.
-For a particular measurement system, which may include radiochemical separation:
D LLD =
E - V - 2.22 - Y -
exp(-Aat)
Wnere:
LLD = the "a priori" lower limit of detection (picoCuries per unit mass or volume),
s = the standard deviation of the background counting rate or of the b counting rate of a blank sample as appropriate (counts per minute),
E = the counting efficiency (counts per disintegration),
V = the sample size (units of mass or volume),
2.22 = the number of disintegrations per minute per picocurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclides (sec 1), and At = the elapsed time between environmental collection, or end of the sample collection period, and time of counting (sec).
Typical values of E, V, Y, and at should be used in the calculation, i
l W SREC 3/4 12-11
DMFI TABLE 4.12-1 (Continued)
TABLE NOTATIONS (Continued)
It should be, recognized that the LLD is defined as an a priori (before the
. fact) limit representing the capability of a-measurement system and not as an a posteriori (after the fact) limit for a particular measurement.
AnaIyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or.
other uncontrollable circumstances may render these LLDs unachievable.
In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Control 6.9.1.3.
l l
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l I W SREC 3/4 12-12 I !
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RADIOLOGICAL ENVIRONMENTAL MONITORING gg 3/4.12.2 LAND USE CENSUS CONTROLS 3.12.2 In accordance with [ plant name] TS 6.8.4.h.2), a Land Use Census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk anima * ,
the nearest residence, and the nearest garden
- of greater than 50 2m (500 ft2) producing broad leaf vegetation. [For elevated releases as defined in Regula-tory Guide 1.111, Revision 1, July 1977, the Land Use Census shall also identify within a distance of 5 km (3 miles) the locations in each of the 16 meteorological sectors of all milk animals and all gardens of greater than 50 m2 producing broad leaf vegetation.]
APPLICABILITY: At all times.
ACTION:
- a. With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Control 4.11.2.3, pursuant to Control 6.9.1.4, identify the new location (s) in the next Semiannual Radioactive Effluent Release Report.
- b. With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently
- being obtained in accordance with Control 3.12.1, add the new location (s) within 30 days to the Radiological Environmental Moni-toring Program given in the ODCM. The sampling location (s), exclud-ing the control station location, having the lowest calculated dose or dose commitment (s), via the same exposure pathwcy, may be deleted from this monitoring program after [0ctober 31] of the year in which this Land Use Census was conducted. Pursuant to Control 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table (s) for the ODCM reflecting the new location (s) with informa-tion supporting the change in sampling locations.
- c. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
i Broad leaf vegetation sampling of at least three different kinds of vegetation l may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted 3/Qs in lieu of the garden census.
Controls for broad leaf vegetation sampling in Table 3.12-1, Part 4.c., shall I
be followed, including analysis of control samples.
W SREC 3/4 12-13 w_____-____________
\
I 4
RADIOLOGICAL ENVIRONMENT 6 MONITORING'-
SURVEILLANCE REQUIREMENTS e
- 4.12.2 ' The i.and Use Census - shall be conducted during the growing season at -
'least once. per 12 months using that information that will. provide the best results, such ~ as .by a door-to-door survey, aerial survey, or by consulting .
local agriculture authorities. The results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report pursuant to Control 6.9.1.3.
i L
W SREC 3/4 12-14 L- _ - - - .- .-. _ _ _ _ - - - - - - _ - - _ - - _____ _ __ ___ _ _ _ _ _ _ _ _ _ __ _ _ _ _
RADIOLOGICAL ENVIRONMENTAL MONITORING drift 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM CONTROLS-3.12.3 In'accordance With [ plant name] TS 6.8.4.h.3), analyses shall be
- performed on all radioactive materials, supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission, that correspond to samples required by Table 3.12-1.
APPLICABILITY: At all times.
ACTION:
- a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Control 6.9.1.3.
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.12.3 The Interlaboratory Comparison Program shall be described in the ODCM.
A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental
' Operating Report pursuant to Control 6.9.1.3.
S W SREC 3/4 12-15
.. n BASES FOR SECTIONS 3.0 AN0'4.0-CONTROLS AND SURVEILLANCE REQUIREMENTS NOTE The BASES contained in succeedin0 pages summarize the reasons for the Controls in Sections 3.0 and 4.0, but are not part of these Controls.
t O
e B 3/4 0-0
DRAFT INSTRUMENTATION BASES 3/4.3.3.10 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent ins +.rumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY-and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
3/4.3.3.11 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous efflu-ents during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
W SREC B 3/4 3-6
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3/4.11 RADI0 ACTIVE EFFLUENTS BASES .
{
3/4.11.1 LIQUID EFFLUENTS l
3/4.11.1.1 CONCENTRATION I f
This control is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within: (1) the Section II.A design objectives of 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. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
This control applies to the release of radioactive materials in liquid effluents from all units at the site.
The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs).
Detailed discussion of the LLD, and other detection limits can be found in Currie, L. A. , " Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements,"
NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300.
3/4.11.1.2 DOSE This control is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix 1, 10 CFR Part 50. The Control implements the guides set forth in Section II.A of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radio- ~~
active material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclides concentrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose calculation methodology and parameters in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calcu-lational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substan-tially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of ySREC B 3/4 11-1
RADIOACTIVE EFFLUENTS BASES DDSE (Continued)
Reactor E1Duents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October.1977 and Regulatory Guide 1.113. " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases
'for the Purpose of Implementing Appendix I," April 1977.
. This control applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared Radwaste Systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system.
3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This control implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified
~1imits governing the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set.
forth in Section II.A of Appendix 1, 10 CFR Part 50 for liquid effluents.
This control applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared Radwaste Systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system.
4 i
ySREC B 3/4 lb2 L _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __________________i
l'. .
RADI0 ACTIVE EFFLUENTS OWI BASES 3/4.11.2 GASEOUS EFFLUENTS 3/4.11.2.1 DOSE RATE This control is proviced to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS.
The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column I. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/ year to the whole body or to less than or equal to 3000 mrems/ year to the skin.
These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than er equal to 1500 mrems/ year.
This control applies to the release of radioactive materials in gaseous effluents from all units at the site.
The required detection capabilities for radioactive material in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs).
Detailed discussion of the LLD, and other detection limits can be found in Currie, L. A. , " Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements,"
NUREG/CR-4007 (September 1984), and in the HASL, Procedures Manual, HASL-300.
3/4.11.2.2 DOSE - NOBLE GASES This control is provided to implement the requirements of Secticns 11.9 III.A and IV.A of Appendix I, 10 CFR Part 50. The control implements the guides set forth in Section I.B of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radio-active material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation
}{SREC _ B 3/4 11-3
DRAFT RADI0 ACTIVE EFFLUENTS 4ASES DOSE-NOBLE GASES (Continued) methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents
. are consistent with the methodology provided in Regulatory Guide 1.109,
" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents I for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"
Revision I, October 1977 and Regulatory Guide '1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases frcm Light-Water Cooled Reactors," Revision 1, July 1977. The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.
This contiol applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared Rsdwaste Treatment Systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.
3/4.11.2.3 DOSE - IODINE-131, 10 DINE-133, TRITIUM, AND RADI0 ACTIVE MATERIAL IN PARTICULATE FORM This contro1 ~is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Controls are the guides set forth in Section II.C of Appendix I. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth'in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section III. A of Appendix I that conformance with'the guides of Appendix I be shown by calcu-lational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substan-tially underestimated. The ODCM calculational methodology and parameters for e m ulating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 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,"
Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate controls for Iodine-131 Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radio sclide pathways to man in the W SREC B 3/4 11-4
( DRAFT RADIOACTIVE EFFLUENTS BASES (
DOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADI0 ACTIVE MATERIAL IN PARTICULATE FORM (Continued) areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of the calculations were: (1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure of man.
This control applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.
3/4.11.2.4 GASEOUS RADWASTE TREATMENT SYSTEM The OPERABILITY of the WASTE GAS HOLDUP SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This control implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section II.D of Appendix I to'10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.
This control applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.
W SREC B 3/4 11-5
- . - -' ? u
' RADIOACTIVE EFFLUENTS BASES
'3/4.11.2.5 NOT USED' P
3/4 11.2.6 NOT USED
~
3/4.11.3 'NOT'USED 3/4.11.4 TOTAL DOSE
. This control is provided to meet the dose limitations of 10 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The control requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioactivity and to radiation from
. uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.
For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR
.Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units W SREC- B 3/4 11-6
._m-___._..__ m . _ . . _ _
DPdH l RADI0 ACTIVE EFFLUENTS l BASES TOTAL DOSE (Continued) i (including outside storage tanks, etc.) are kept small. The Special Report 1 will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits.
For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER of the PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is com-pleted. The variance only relates to the limits of 40 CFR Part 190, and does
,not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Controls 3.11.1.1 and 3.11.2.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.
e HSREC B 3/4 11-7
1
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4 . q d l Li i 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING
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3 BASES 3/4.12.1 MONITORING PROGRAM The Radiological Environmental Monitoring Program required by this 1 control provides representative measurements of radiation and of radioactive 1 materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposure of MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby 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 the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring, Revision 1, November 1979. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience.
The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table 4.12-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measure-ment system and not as an a posteriori (after the fact) limit for a particular measurement.
Detailed discussion of the LLD, and other detection limits, can be found in Currie, L. A. , " Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements,"
NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300.
3/4.12.2 LAND USE CENSUS This control is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that~ modifications to the Radiological Environmental Monitoring Program given in the ODCM are made if required by the results of this census. The best information from the door-to-door survey, frem aerial survey or from consulting with local agri-cultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50 m2 provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden I of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child.
To determine this minimum garden size, the following assumptions were made:
(1) 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m2, i
1 y SREC B 3/4 12-1
\ . _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _
DRWI RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50.
e W SREC B 3/4 12-2
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j L SECTION 5.0 DESIGN FEATURES 1
ySREC 5-0 l
. _ _ - - - - _ _ _ _ _ _ _ _ _ _ . i
4.
f
' 5. 0 DESIGN FEATURES-5.1 SITE i
MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as defini-i tion of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to l MEMBERS OF THE PUBLIC, shall be as shown in Figure [5.1-3].
The definition of UNRESTRICTED AREA used in implementing these Controls has-been expanded over that in 10 CFR 20.3(a)(17). -The UNRESTRICTED AREA boundary may coincide with the Exclusion (fenced) Area boundary, as defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas over water bedies. The concept of UNRESTRICTED AREAS, established at or beyond the SITE BOUNDARY, is utilized in the Controls to keep levels of radioactive materials in liquid and gaseous effluents as low as is reasonably achievable, pursuant
.to 10 CFR 50.36a.
i l
W SREC 5-1 1
- - 1 This figure shall consic of a map of the site area showing the SITE BOUNDARY and locating points within the SITE BOUNDARY where radioactive gaseous and liquid effluents are released, as well as where radioactive liquid effluents leave the site. If onsite areas sub-ject to radioactive materials in gaseous or liquid effluents are utilized by the public for recreational or other purposes, these areas shall be outlined on the map and identified by occupancy control (if any).
The figure shall be sufficiently detailed to allow identification of structures and release point locations and elevations, as well'as definition of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE-PUBLIC. The map scale shall be on the. order of 2-3"/ mile. See NUREG-0133 for additional guidance.
1 l
l I
i 1
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4 i
FIGURE 5.1-3 l
UNRESTRICTED AREA AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND LIQUID EFFLUENTS
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W SREC 5-4
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DRA!T l
SECTION 6.0 ADMIi4ISTRATIVE CONTROLS O
W SREC 6-0
.. _ _ _ _ _ _ _ . ________ _ ___ _ _ _ _ a
J' ' 4
- ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT **
J5.9.1.3 Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality and shall include copies of reports of the preoperational Radiological Environmental Monitoring Program of the unit for at least two years prior to initial criticality.
The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, snd an analysis of trends of the results of the radiological env!ronmental surveillance activities for the report period,
- A single submittal may be made for a multiple unit station.
W SREC G-17
l' DRAFT ADMINISTRATIVE CONTR08.5 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (Continued) including a comparison with preoperational studies and with operational i controls, as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the 1 enviremment. The reports shall also include the results of the Land Use Census required by Control 3.12.2.
The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the Offsite Dose Calculation Manual, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some indivi-dual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
The reports shall also include the following: a summary description of the Radiological Environmental Monitoring Program; at least two legible maps
- covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program and the corrective action taken if the specified program is not being performed as required by Control 3.12.3; reasons for not conducting the Radiological Environmental Monitoring Program as required by control 3.12.1, and discussion of all deviations from the sampling schedule of Table 3.12-1; discussion of environmental sample measure-ments that exceed the reporting levels of Table 3.12-2 but are not the result of plant effluents, pursuant to ACTION b. of Control 3.12.1; and discussion of all. analyses in which the LLD required by Table 4.12-1 was not achievable.
SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT **
6.9.1.4 Routine Semiannual Radioactive Effluent Release Reports covering the _
operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial crit.icality.
The Semiannual Radioactive Effluent Re19ase 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," Revision 1, June 1974, with data
- 0ne map shall cover stations near the SITE BOUNDARY; a second sr.all include the more distant stations.
- A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
W SREC 6-18
ADMINISTRATIVE CONTROLS DE i SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued) summarized on a quarterly basis following the format of Appendix B thereof.
For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories: class of solid wastes (as defined by 10 CFR Part 61), type of container (e.g., LSA, Type A, Type B, large Quantity) and SOLIDIFICATION agent or absorbent (e.g. , cement, urea formaldehyde).
The Semiannual Radioactive Effluent Release Report to be submitted within 60 days af ter January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability." This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit er station during the previous calendar year. This same report shali also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure [5.1-3]) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time, and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous efflu-eids, as deterniined by sanpling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (DDCM).
The Semiannual Radioactive Effluent Release Report to be submitted within 60 days af ter January 1 of each year shall also include an assessment of radia-tion doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary ef flu-ent pathways and direct radiation, for the previous calendar year to show con-formance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribu-tion from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, 0:tober 1977. -
The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.
The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) l
- In lieu of submission with the Semiannual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.
6-19 M SREC
? . .
L DRiff
' ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), pursuant to Control 6.14, as well as any major change to Liquid, Gaseous, or Solid Rac' waste Treatment Systems. It shall also include a. listing of new locations for dose calcu-lations and/or environmental monitoring identified by the Land Use Census pursuant to Control 3.12.2.
The Semiannual Radioactive Effluent Release Reports shall also include the following: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Control 3.3.3.10 or 3.3.3.11, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the control limits.
W SREC 6-20 l
- - - - - - - - - - - - - - - - - - - = - =
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DRAFT
-ADMINISTRATIVE CONTR0LS~
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4 6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM)
. Changes to the ODCM:
. a. Shall be' documented and records of reviews performed shall be retained as required by Specification 6.10.30. This documentation shall contain:
- 1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and 2), A determination that the change will-maintain the level of radioactive effluent control required by 10 CFR 20.1'4, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or. reliability of effluent,.
dose, or setpoint calculations.
E
- b.- Shall become' effective after review and acceptance by the [URG) and I. the approval of the Plant Manager,
- c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as part of or concurrent with the-Semiannual Radioactive Effluent Release Report for-the period of the l report in which any change to the ODCM was made. Each change shall )
be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented. .
)
W SREC 6-24
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b7 - - - --_ ___
e s APPENDIX A Radiological Assessment Branch Technical Position, Revision 1, November 1979
.. 1 Revision 1 November 1979 Branch Technical Positirp Background o Regulatory Guide 4.8, Environmental Technical Specifications for Nuclear Power Plants, issued for comment in December 1975, is being revised based on comments received. The Radiological Assessment Branch issued a Branch Position on the radiological portion of the environmental monitoring program in March,1978.
The position was formulated by an NRC working group which considered comments received after the issuance of the Regulatory Guide 4.8. This is Revision 1 of that Branch Position paper. The changes are marked by a vertical line in the right margin. The most significant change is the increase in direct radiation measurement stations.
20 CFR Parts 20 and 50 require that radiological environmental monitoring programs be established to provide data on measurable levels of radiation and radioactive materials in the site environs. In addition, Appendix ! to 10 CFR Part 50 requires that the relationship batween quantities of radioactive material released in effluents during normal operation, including anticipated operational occurrences, and resultant radiation doses to individuals frem principals pathways of exposure be evaluated. These pngrams should be con-ducted to verify the effectiveness of in plant measures qsed for controlling the release of radioactive materials. Surveillance shoub! be established to identify changer, in tne use cf unrestricted areas (e.g. , ft. agricultrual purposes) to provide a basis for modifications in the monitoring programs for evaluating doses to individuals from principal pathways of exposure. NRC Regulatory Guide 4.1, Rev.1, " Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants," provides an acceptable basis for the design of programs to monitor levels of radiation and radioactivity in the station environs.
This position sets forth an example of an' acceptable minimum radiological monitoring program. Local site characteristics must be examined to determine if pathways not covered by this guide may significantly contribute to an individual's dose and should be included in the sampling program.
I 1
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2 4
AN ACCEPTABLE RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Program Requirements Environmental samples shall be collected and analyzed according to Table 1 at locations shown in Figure 1.1 Analytical techniques used shall be such that the detection capabilities in Table 2 are achieved.
The results of the radiological envi nnmental monitoring are intended to supf ement the results of the radiological effluent monitoring by verifying thathe measurable concentrations of radioactive materials and levels of radiation are net hi.2her than expected on the basis of the affluent measure-ments and modeling of the environmental exposure pathways. Thus, the specified environmental monitoring program provides measurements of radiation.and of radio-active materials in those exposure pathways and for those radionuclides which lead to the highest r.otential radiation exposures of individuals resulting from the station operation. The initial radiological environmental monitoring program should be conducted for the first three years of wmmercial operation (or other period corresponding to a maximum burnup in the initial core cycle). Following this period, program changes may be proposed based on operational experience.
The specified detection capabilities are state-of-the-art for routine environ-mental measurements in industrial laboratories.
Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimans are unobtainable due to sampling equipment malfection, every effc. t shall be made to complete corrective action prior to the end of the next sanpling period. All deviations from the sampling schedule shall be documented in the annual report. __
The laboratories of the licensee and licensee's contractors which perform analyses shall participate in the Environmental Protection Agency's (EPA's)
Environmental Radioactivity Laboratory Intercomparisons Studies (Crosscheck)
Program or equivalent program. This participation shall include all of the determinations (sample medium-radionuclides combination) that are offered by EPA and that also are included in the monitoring program. The results of analysis of these crosscheck samples shall be included in the annual report. I The participants in the EPA crosscheck program may provide their EPA program code so that the NRC can review the EPA's participant data directly in lieu of submission in the annual report.
2 It may be necessary to require special studies on a case-by-case and site specific basis to establish the* relationship between quantities of radioactive material released in effluents, the concentrations in environmental media, and the resultant doses for important pathways.
r s 3
If the results of a determination in the EPA crosscheck program (or equivalent
. program) are outside the specified control limits, the laboratory shall inves-tigate the cause of the problem and take steps to correct it. The results cf this investigation and corrective action shall be included in the annual report.
The requirement for the participation in the EPA crosscheck program, or similar program, is based on the need for independent checks on the precision and accurso of the measurements of radioactive material in environmental sample '
matri, > as part of the quality assurance program for environmental monitoring in oro.c to demonstrate that the results are reasonably valid.
A census shall be conducted annually during the growing season to determine the location of the nearest milk animal and nearest garden greater than 50 square meters (500 sq. ft.) producing broad leaf vegetation in each of the 16 meteorological sectors within a distance of 8 km (5 miles).8 For elevated releases as defined in Regulatory Guide 1.111 Rev. 1., the census shall also
, identify the locations of all milk animals, and gardens greater than 50 square meters producing broad leaTvegetation out to a distance of 5 km. (3 miles) for each radial sector.
If it is learned from this census that the milk animals or gardens are present at a location wnich yields a calculated thyroid dose greater than those previously sampled, or if the census resul s in changes in the location used in the radioactive effluent technical specifications for dose calculations, a written report shall be submitted to thc Director of Operating Reactors, NRR (with a copy to the Director of the NRC Regional Office) within 30 days identifying the new location (distance and direction). Milk anish1 or garden locations resulting in higher calculated doses shall be added to the surveillance program as soon as practicable.
The sampling location (excluding the control sample location) having the lowest calculated dose may then be dropped from the surveillance program at the end of the grazing or growing season during which the census was con-ducted. Any location from whien milk can no longer be obtained may be dropped from the surveillance program af ter notifying the NRC in writing that they are no longer obtainable at that location. The results of the land-use census shall be reported in the annual report.
The census of milk animals and gardens producing broad leaf vegetation is based on the requirement in Appendix I of 10 CFR Part 50 to " Identify changes in the use of unrestricted areas (e.g. , for agricultural purposes) to permit modifications in monitoring programs for evaluating doses to individuals from principal pathways of exposure. The consumption of milk from animals grazing on contaminated pasture and of la.fy vegetation conte'ninated by airborne z
Broad leaf vegetat'en sampling r.sy be perfomed at the site boundary in a sector with the highest D/Q in i.eu of the garden census.
l n
' I ,
4 radiciodine is a major potential source of exposure. Samples from milk animals are considered a better indicator of radiciodine in the environment than vegetation. If the census reveals milk animals are not present or are unavailab~e for sampling, then vegetation must be sampled.
The 50 square meter garden, considering 20% used for growing broad leaf vegetation 2
(i.e., similar to lettuce and cabbage), and a vegetation yield of 2 kg/m ,
will produce the 26 kg/yr assumed in Regulatory Guide 1.109 Rev 1., for child consumption of leafy vegetation. The option to consider the garden to be broad leaf vegetation at the site Madary in a sector with the highest 0/Q should be conservative and that location may be used to calculate doses due to radioactive effluent releases in place of the actual locations which would be determined by the census. This option does not apply to plants with elevated re12ases as defined in Regulatory Guide 1.111, Rev.1.
The increase in the number of direct radiation stations is to better characterize the individual exposure (mrem) and population exposure (man-rem) in accordance with Criterion 64 - Monitoring radioactivity releases, of 10 CFR Part 50, Appendix A. The NRC will place a similar amount of ste.tions in the area between the two rings designated in Table 1.
NOTE Guidance on the subjects contained on pages 4 through 16 of the Radiological Assessment Branch Technical Position (RAB-BTP) has been modified and upgraded based on operating experience since Revision 1 was published in 1979. The current staff guidance for the following items has been incorporated in the Section 3/4-12 and Section 6 Controls of NUREG-1301 and 1302.
- Reporting Requirement
- Table 1: Operational Radiological Environmental Monitoring Report
- Table 2: Detection Capabilities for Environmental Sample Analysis
- Table 4: Reporting !.evels for Radioactivity Concentrations in Environmen-tal Samples The following items remain unchanged:
- Footnote to Table 1 nn page 10
- Table 3 of page 14 Figure 1 of page 16
/ ' f. I Pages 5, 6, 7, 8, 9, 11, 12, 13, 15 I
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The above pages have been superceded by text and tables in NUREG-1301 and 1302.
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APPENDIX'B
" Appendix B - General Contents ^of the Offsite Dose Calculation Manual (0DCM) (Revision 1, February 1979)" to the paper authored by C. A. Willis and F. J. Congel, " Status of NRC Radiological Effluent Technical Specification Activities" presented at the Atomic Industrial Forum Conference on NEPA and Nuclear Regulation, October 4-7, 1981, Wa shi ngton , ' D.C.
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7 APPENDIX B GENERAL CONTENTS OF THE OFFSITE DOSE CALCULATION MANUAL (ODCM*)
(Rev.1, February 1979)
Section 1 - Set Points Provide the equations and methodology to be used at the station or unit for each alarm and trip set point on each effluent release point according to the Specifications 3.3.3.8 and 3.3.3.9. The instrumentation for each alarm and trip set point, including radiation monitoring and samp' sing systems and ef fluent control features, should be identified by reference to the FSAR (or Final Hazard Summary). This information should be consistent with the recommendations of Section I of Standard Review Plan 11.5, NUREG-75/087, (Revision 1). If the alarm and/cr trip set point value is variable, provide the equation to determine the set point value to be used, based on actual release conditions, that will assure that the Specification is met at each release point; and provide the value to be used when releases are not in progress. If dilution or dispersion is used, state the onsite equipment and measurement method used during release, the site related parameters and the set points used to assure that the Specification is met at each release point. The fixed and variable set points should consider the radioactive effluent to have a radionuclides distribution represented by normal and anticipated operational occurrences.
Section 2 - Liquid Effluent Concentration Provide the equations and methodology to be used at the station or unit for each liquid release point according to the Specification 3.11.1.1. For systems with continuous or batch releases, and for systems designed to monitor and control both continuous and batch releases, provide the assump-tions and parameters to be used to compare the output of the monitor with the liquid concentration specified. State the limitations for combined discharges to the same release point. In addition, describe the method and assumptions for obtaining representative samples from each batch and use of previous post-release analyses or composite sample analyses to meet the Specification.
Section 3 - Gaseous Effluent Dose Rate Provide the equations and methodology to be used at the station or unit for each gaseous release point according to Specification 3.11.2.1. Consider the various pathways, release point elevations, site related parameters and radionuclides contribution to the dose impact limitation. Provide the
- The format for the ODCM is left up to the licensee and may be simplified by tables and grid printout. Each page should be numbered and indicate the facility approval and ef fective date. .
j 1
1 dose factors to be used for the identified radionuclides released. Provide the annual average dispersion values (X/Q and D/Q), the site specific para-meters and release point elevations.
Section 4 - Liouid Effluent Dose Provide the equations and methodology to be used at the station or unit for each liquid release point according to the dose objectives given in Speci-fication 3.11.1.2. The section should describe how the dose contributions are to be calculated for the various pathways and release points, the equa-tions and assumptions to be used, the site specific parameters to be measured and used, the receptor location by direction and distance, and the method of estimating and updating cumulative doses due to liquid releases. The dose factors, pathway transfer f actors, pathway usage factors, and dilution fac-tors for the points of pathway origin, etc., should be given, as well as receptor age group, water and food consumption rate and other factors assumed or measured. Provide the method of determining the dilution factor at the discharge during any liquid effluent release and any site specific parameters used in these determinations.
Section 5 - Gaseous Effluent Dose Provide the equations and methodology to be used at the station or unit for each gaseous release point according to the dose objectives given in Specifications 3.11.2.2 and 3.11.2.3. The section should describe how the dose contributions are to be calculated for the various pathways and release points, the equations and assumptions to be used, the site specific parameters to be measured and used, the receptor location by direction and distance, and the method to be used for estimating and updating cumulative doses due to gaseous releases. The location, direction and distance to the nearest resi-dence, cow, goat, meat animal, garden, etc., should be given, as well as receptor age group, crop yield, grazing time and other factors assumed or measured. Provide the method of determining dispersion values (X/Q and D/Q) for releases and any site specific parameters and release point elevations used 'n these determinations.
Section 6 - Projected Doses For liquid and gaseous radwaste treatment systems, provide the method of projecting doses due to effluent releases for the normal and alternate pathways of treatment according to the specifications, describing the com-ponents and subsystems to be used.
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- 3-Section 7 - Operability of Equipment Provide a. flow diagram (s) defining the treatment paths and the components of the radioactive liquid, gaseous and solid waste management systems that are
. to be maintained and 'used, pursuant to 10 CFR 50.36a, to meet Technical Specifications 3.11.1.3, 3.11.2.4 and 3.11.3.1. Subcomponents of packaged equipment can be identified by a list. For operating reactors whose con-struction permit applications were filed prior to January 2,1971, the flow diagram (s) shall be consistent with the infomation provided in confonnance with Section V.B.1 of Appendix I to 10 CFR Part 50. For OL applications whose construction permits were fiied af ter January 2,1971, the flow )
diagram (s) shall be consistent with the infonnation provided in Chapter 11 of the Final Safety Analysis Report (FSAR) or amendments thereto.
Section 8 - Sample Locations ,
1 Provide a map of the Radiological Environmental Monitoring Sample Locations i indicating the numbered sampling locations given in Table 3.12-1. Further clarification on these numbered sampling locations can be provided by a list, indicating the direction and distance from the center of the building com-plex of the unit or station, and may include a descriptive name for identi-fication purposes.
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APPENDIX C GENERIC LETTER'89-01 IMPLEMENTAT]DN.0F PROGRAMMATIC CONTROLS FOR RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS IN THE ADMINISTRATIVE CONTROLS SECTION OF THE TECHNICAL SPECIFICATIONS AND THE RELOCATION OF PROCEDURAL DE OF RETS TO THE OFFSITE DOSE CALCULATION MANUAL OR TO THE PROCESS CONTROL PROGRAM 6
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p mus UNITED STATES
..[ g NUCLEAR RECULATORY COMMISSION 7, CAWWGTON, D. C. 205$$
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/ January 31, 1989 l
TO ALL POWER REACTOR LICENSE.S AND APPLICANTS
SUBJECT:
IMPLEMENTATION OF PROGRAMMATIC CONTROLS FOR RADIOLOGICAL EFFLUENT- l TECHNICAL SPECIFICATIONS IN THE ADMINISTRATIVE CONTROLS SECTION OF I
THE TECHNICAL SPECIFICATIONS AND THE RELOCATION OF PROCEDURAL DETAILS 0F FETS TO THE OFFSITE DOSE CALCULATION MANUAL OR TO THE PROCESS I CONTROL PROGRAM (GENERIC LETTER 89-01) l The NRC staff has examined the contents of the Radiological Effluent Technical Specifications (RETS) in relation to the Conunission's Interim Policy Statement on Technical Specification Improvements. The staff has detemined that pro-grancatic the Technicalcontrols can be imp Specifications (lemmt:d TS) to 4 the Administrative satisfy existing Controls section of regulatory reoutrements for RETS. At the same tire, the procedural details of the current TS on radio-active affkents and radiological environmental monitoring can be relocated to theOffsiteDoseCalculationManual(ODCM). Likewise, the procedural details of the current TS on solid radioactive wastes can be relocated to the Process Control Program (PCP). These actions simplify the RETS, meet the regulatory reoutrements for radioactive effluents and radiological environmental monitor- i ing, and are provided as a line-item improvement of the TS, consistent with the goals of the Policy Statement. ,
New programmatic controls for radioactive effluents and radiological environ-mental monitoring are incorporated in the TS to confom to the regulatory requirements of 10 CFR 20.106, 40 CFR Part 190, 10 CFR S0.36a, and Appendix I to 10 CFR Part 50. Existing prograsunatic recuiresents for the PCP are being retained in the TS. The procedural details included in licensees' present TS cn radioactive effluents, solid radioactive wastes, environmental monitoring, and associated reporting reautrements will be relocated to the ODCM or PCP as appropriate. Licensees will handle future changes to these procedural details -
in the ODCM and the PCP under the administrative controls for chenges to the ODCH or PCP. Finally, the definitions of the ODCM and PCP are updated to reflect these changes.
Enclosure 1 provides guidance for the preparation of a license amendment re-ouest to implement these alternatives for RETS. Enclosure 2 provides a list-ing of existing RETS and a description of how each is addressed. Enclosure 3 provides model TS for programmatic controls for RE15 and its associated report-ing reautrements. Finally.. Enclosure 4 provides model specifications for retaining existing requirements for explosive gas monitoring instrumentation requirements that apply on a plant-specific basis. Licensees are encouraged to propose changes to TS that are consistent with the guidance provided in the enclosures. Conforming atendment recuests will be expeditiously reviewed by
f' .
f t
January 31,1989.
- deneric Lctter 89-01 2 l.
1
-the NRC Project Manager for the facility. Proposed amendments that deviate from this guidance will require a longer, more detailed review. Please contact j l' the' appropriate Project Manager if you have questions on this matter. J n
Sincercly, j .
. aY a .
4 Acting Associate D ctor for Projects Office of Nuclear Reactor Regulation * *
Enclosures:
1 through 4 as stated 4
82neric Letter 89 01 ENCLOSURE 1 GUIDANCE FOR THE IMPLEMENTATION OF PROGRAMMATIC CONTROLS FOR RETS IN THE ADMINISTRATIVE CONTROLS SECTION OF TECHNICAL SPECIFICATIONS AND THE RELOCATION OF PROCEDURAL DETAILS OF CURRENT RETS TO THE OFFSITE DOSE CALCULATION MANUAL OR PROCESS CONTROL PROGRAM INTRODUCTION This enclosure provides guidance for the preparation of a license amendment request to implement programmatic controls in Technical Specifications (TS) for radioactive effluents and for radiological environmental monitoring con-forming to the applicable regulatory requirements. This will allow the reloca-tion of existing procedural details of the current Radiological Effluent Technical Specifications (RETS) to the Offsite Dose Calculation Manual (0DCM).
Procedural details for solid radioactive wastes will be relocated to the Process Control Prograr, (PCP). A proposed amendment will (1) incorporate pro-grammatic controls in the Administrative Controls section of the TS that sat-isfy the requirements of 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a. and Appendix I to 10 CFR Part 50, (2) relocate the existing procedural details ~ in current specifications involving radioactive effluent monitoring instruments-tion, the control of liquid and gaseous effluents, equipment requirements for liquid and gaseous effluents, radiological environmental monitoring, and radio-logical reporting details from the TS to the ODCM, (3) relocate the definition of solidification and existing procedural details in the current specification on solid r.adioactive wastes to the PCP, (4) simplify the associated reporting requir m ats, (5) simplify the administrative controls for changes to the ODCM and PCP, (6) add record retention requirements for changes to the ODCM and PCP, and (7) update the definitions of the ODCM and PCP consistent with these changes.
The NRC staff's intent in recommending these changes to the TS and the reloca-tion of procedural details of the current RETS to the ODCM and PCP is to ful-fill the goal of the Commission Policy Statement for Tc:hnical Specification Improvements. It-is not the staff's intent to reduce the level of radiological effluent control. Rather, this amendment will provide programmatic controls for RETS consistent with regulatory requirements and allow relocation of the procedural details of current RETS to the ODCM or PCP. Therefore, future changes to these procedural details will be controlled by the controls for changes to the ODCM or PCP included in the Administrative Controls section of the TS. These procedural details are not required to be included in TS by 10 CFR 50.36a.
DISCUSSION Enclosure 2 to Generic Letter 89- provides a summary listing of specifica-tions that are included under the heading of RETS in the Stan d rd Technical Specifications (STS) and their disposition. Most of these specifications will be addressed by programmatic controls in the Administrative Controls section of the TS. Some specifications under the heading of RETS are not covered by the new programmatic controls and will be retained as requirements in the existing plant TS. Examples include requirements for explosive gas monitoring instru-mentation, limitations on the quantity of radioactivity in liquid or gaseous holdup or storage tanks or in the condenser exhaust for BWRs, or limitations on explosive gas mixtures in offgas treatment systems and storage tanks.
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, Ceneric Letter 89- 01 Enclesure 1 I . Licensees with nonstandard TS should follow the guidance provided in Enclo-sure 2 for the disposition of similar requirements in the format of their TS.
( Because solid radioactive wastes are addressed under existing programmatic l controls for the Process Control Program, which is a separate program from the new programmatic controls for liquid and gaseous radioactive effluents, the requirements for solid radioactive wastes and associated solid waste reporting requirements in current TS are included as procedural details that will be relocated to the PCP as part of this line-item improvement of TS. Also, the staff has concluded that records of licensee reviews performed for changes made to the ODCM and PCP should be documented and retained for the duration of the unit operating license. This approach is in lieu of the current requirements that the reasons for changes to the ODCM and PCP be addressed in the Semiannual Effluent Release Report.
The following items are to be included in a lic.nse amendment request to imple-ment these changes. First, the model specifications in Enclosure 3 to Generic Letter 89- should be incorporated into the TS to satisfy the requirements of 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50.
The definitions of the ODCM and PCP should be updated to reflect these changes.
The programmatic and reporting requirements are general in nature and do not contain plant-specific details. Therefore, these changes to the Administrative Controls section of the TS are to replace corresponding requirements in plant 1 TS that address these items. They should be proposed for incorporation into the plant's TS without change in substance to replace existing requirements. !
If necessary, only changes in format should be proposed. If the current TS {
include requirements for explosive gas monitoring instrumentation as part of I the gaseous effluent monitoring instrumentation requirements, these require- i ments should be retained. Enclosure 4 to Generic Letter 89- provides model j specifications for retaining such requirements. i j i
Seconds the procedural details covered in the licensee's current RETS, consist-ing of the limiting conditions for operation, their applicability, remedial actions, surveillance requirements, and the Bases section of the TS for these requirements, are to be relocated to the ODCM or PCP as appropriate and in a manner that ensuras that these details are incorporated in plant operating pro-cedures. The HRC staff does not intend to repeat technical reviews of the re-located procedural details because their consistency with the applicable regula-tory requirements is a matter of record from past NRC reviews of RETS. If licensees make other than editorial changes in the procedural details being transferred to the ODCM, each change should be identified by markings in the !
margin and the requirements of new Specification 6.14a.(1) and (2) followed. I Finally, licensees should confirm in the amendment request that changes for .
relocating the procedural details of current RETS to either the ODCM or PCP l have been prepared in accordance with the proposed changes to the Administra- l tive Controls section of the TS so that they may be implemented immediately ;
upon issuance of the proposed amtndment. A complete and legible copy of the i revised ODCM should be forwarded with the amendment request for NRC use as a reference. The NRC staff will not concur in or approve the revised ODCH.
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-' Ceneric Letter 89-01 Enc 1:sure 1 Licensees should refer to " Generic Letter 89-
- in the Sub.iect line of license amendment reauests implementing the guidance of this Generic Letter. This will facilitate the staff's tracking of licensees' responses to this Generic Letter.
1
SUMMARY
! The license amendment reauest for the line-item improvements of the TS relative to the RETS will entail (1) the incorporation of programmatic controls for L- radioactive effluents and radiological environmental monitoring in the Admin-istrative Controls section of the TS, (2) incorporatation of the procedural details of the current RETS in the ODCM or PCP as appropriate, and (3) confirm-ation that the guidance of this Generic Letter has been followed.
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- - Generic Lett r 89-01 Enclosure 3 l
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TECHNICAL SPECIFICATIONS TO BE REVISED 1.17 DEFINITIONS: OFFSITE DOSE CALCULATION MANUAL ,
i 1.22 -DEFINITIONS: PROCESS CONTROL PROGRAM J
6.8.4 g. PROCEDURES AND PROGRAMS: RADI0 ACTIVE EFFLUENT CONTROLS 6.8.4 h. PROCEDURES AND FROGRAMS: RADIOLOGICAL ENVIRONMENTAL MONITORING 6.9.1.3 REPORTING REQUIREMENTS: ANNUAL RADIOLOGICAL ENVIRONMENTAL i OPERATING REPORT 6.9.1.4 REPORTING REQUIREMENTS: SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 6.10 RECORD RETENTION 6.13 PROCESS CONTROL PROGRAM (PCP) 6.14 0FFSITE DDSE CALCULATION MANUAL (ODCM)
MCDEL TECHNICAL SPECIFICATION REVISIONS (To supplement or replace existing specifications) 1.0 DEFINITIONS OFFSITE DOSE CALCULATION MANUAL 1.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and licuid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Fro-grams required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semi-annual Radioactive Effluent Release Reports required by Specifications 6.9.1.3 and 6.9.1.4.
1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that process-ing and packaDi ng of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as
-to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial gtound requirements, and other requirements governing the disposal of solid radioactive waste. -
Generic Letter 89- 01 Enclosure 3 6.0- ADMINISTRATIVE CONTROLS 6.8 PROCEDURES AND PROGRAMS 6.8.4 The following programs shall be established, implemented, and mainte.ined:
- g. Radioactive Effluent Controls Program A program shall be provided confoming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM,.
(2) chall be impicmented by operating procedures, and (3) shall in-clude remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
- 1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and set-point determination in accordance with the methodology in the ODCM,
- 2) Limitations on the concentrator,ns of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix 8 Table II, Column 2,
- 3) . Monitoring, savoling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM,
- 4) Limitations on the annoal and quarterly doses or dose comitment to a MEMBER OF THE PUBLIC from radioactive materials in Epid effluents released from each unit to UNRESTRICTED AREAS onfom-ing to Appendix I to 10 CFR Part 50,
- 5) Determination of cumulative and projseted dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days,
- 6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are.used to reduce releases of radio-activity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50, L 7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the SITE BOUNDARY confoming to the doses associated with 10 CFR Part 20, Appendix B, Table II, Column 1,
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l 1.J Seneric' Letter 8941. - 3'- Enclosure 3 1
ADMINISTRATIVE CONTROLS
{ ' 6.8.4 g. ' Radioactive Effluent Controls Program (Cont.)
'8) Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to' Appendix I to 10 CFR Part 50,
- 9) Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radio-nuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY conforming to Appendix-I to 10 CFR Part 50,
, 10) Limitations on venting and purging of the Mark II containment through the Standoy Gas Treatment System to s,aintain releases as low as reasonably achievable (BWRs w/ Mark II containments),
and
- 11) Limitations on the annual dose or dose commitment to any MEMBER
'0F THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.
- h. Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radio-nuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent meitering program and modeling of environmental expo-sure pathwayr (he gram shall (1) be contained in the ODCM, (2) conform to the gu nce of Appendix I to 10 CFR Part 50, and (3) include the followin . -
- 1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the method-
. ology and parameters in the ODCM,
- 2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifica-tions to the monitoring program are made if required' by the results of this census, and
- 3) Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as pa7t of the quality assurance pro-gram for environmental monitoring.
tr- C'eneric Letter 89901 Enclosure 3 ADMINISTRATIVE CONTROLS 6.9 REPORTING REQUIREMENT _S ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- 6.9.1.3 The Annual Radiological Environmental Operating Report covering the operatinn of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpreta-tions, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2 IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.
SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT **
6.S.I.4 The Semiannual Radioactive Effluent Release Report cover'ing the oper-ation of the unit during the previous 6 months of operation shall be submitted within 60 days af ter January 1 and July 1 of each year. The report shall in-clude a summary of the quantities of radioactive liquid and gaseous effluents and solid waste reletsed from the unit. The material provided shall be
. (1) consistent with the objectives outlined in the ODCM and PCP and (2) in con-formance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.
6.10 RECORD RETENTION 6.10.3 The following records shall be retained for the duration of the unit Operating License:
- o. Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.
6.13 PROCESS CONTROL PROGRAM (PCP)
Changes to the PCP:
- a. Shall be documented and records of reviews performed shall be retain-ed as required by Specification 6.10.30. This documentation shall conta in:
- 1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and
- A single submittal may o* e made for a multi-unit station.
- A single submittal may be made for a multi-unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit. ,
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e y i J Generic LGtter 89-01 ,
Enclosure 3 ADMINISTRATIVE CONTROLS ,
'6.13' PROCESS CONTROL PROGRAM'(PCP) (ront.)
- 2) A determination that the change will maintain the overall con-formance of the' solidified waste product to existing require-ments of Federal, State, or'other applicable regulations.
- b. N 11 become effective after review and acceptance by the [URG) and
- the approval of the Plant Manager.
6.14- 0FFSITE DOSE CALCULATION MANUAL (ODCM)
Changes to the DDCM:
- a. .Shall be' documented and records of reviews performed shall be retain-ed as required by Specification 6.10.30. This documentation shall i contain:
- 1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying'the change (s) and 2)' =A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adver tly' impact the accuracy or reliability of effluent, dose, or setpoint calculations.
- b. Shall become effective after review and acceptance by the [URG) and the approval of the Plant Manager.
. c. Shall be submitted to the Commission in the form of a complete.
legible copy of the entire ODCM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report.for the period of the report in which any change to the ODCH was made. Eact change shall be identified by markings in the margin of the affected pages, _
clearly indicating the area of the page that was changed, and shall indicate the date (e.g. , month / year) the change was implemented.
. _ _ _ . _ _ . _ _ _ .-_m.__m_ _._ _ _ _ _ _ _ .- __.__.__m-.__m. _- _..- - _____.__m__.______-__m
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, Generic Letter 89- 01 Enclosure 4 j MODIFICATION OF THE SPECIFICATION FOR RAD 10 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION TO RETAIN REQUIREMENTS FOR EXPLOSIVE GAS MONITORING INSTRUMENTATION INSTRUMENTATION EXPLOSIVE RABf6AEiiVE GASE885-EFFtBENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION explosive 3.3.3.11 The radienetive gaseens-efficent monitoring instrumentation channels shown in Table 3.3-13 shall be OP2RABLE with their Alarr/ Trip Setpoints set to ensure that the limits of Specifications S:ll:Eri-and 3.11.2.5 are not '
exceeded. The- Al arm /T ri p-S e tpoi nts- o f- the s e- c ha nnei s -me e ti ng-Sp e ci fi c ati on S:ll: E r i- s h ai l- b e- d e t e r mi n e d- a nd- a dj u s t e d-i n- a c c o rd a nc e-wi th- the- me tho d oi e gy and parameters-in-the-6BEM:
APPLICABILITY: As shown in Table 3.3-13 ACTION:
explosive
- a. With an radioactive gaseens-effinent monitering instrumentation channeT Alarm / Trip Setpoint less conservative than required by the above specification--immediately-sespend-the-teiease-of-radiesetive gaseens-efficents-menitored-by-the-effected-channei--er declare the channel inoperable and take the ACTION shown in Table 3.3-13.
explosive
- b. With less than the minimum number of radioactive gaseens-efficent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful explain-in-the-next-Semi-annuai-Radienetive-Efficent-Release-Report prepare and submit a Special Report to the Commission pursuant to Specification 6-9:1:4 6.9.2 to explain why this inoperability was not corrected ir. a timely manner.
- c. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS explosive 4.3.3.11 Each radienetive gaseens-efficent monitoring instrumentation channel shall be demon,trated OPERABLE by performance of the CHANNEL CHECK, 508REE EHEEK- CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-9.
Sample STS 3/4 3-(n)
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d Sntn o o Sntn ono) rurl S e Aiso r r Aiso ddda N s Grhi d d Grhi yeyu u ots y y ots I
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' C .' *i i'-- TABtr 3l3;13' Ttenfinu'e .- T_~ M S 1 -
' 'O' (Not used) c* During WASTE GAS HOLDUP SYSTEM operation.
l D ACTION STATEMENTS f
ACTION 45 - (Not used)-
' ACTION 46 - (Not used)
ACTION 47 - (Not used)
ACTION 48 - (Not used)
ACTION O - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of this WASTE CAS BOLDUP SYSTEM may continue provided grab samples are collected at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and analyzed within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
ACTION 50 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, operation of this system may continue provided grab samples are taken and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With both channels inoperable, operation may continue provided greb samples are taken and analyzed at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during degassing operations and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during other operations.
ACTION 51 -~ (Not used)
ACTION 52 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend oxygen supply to'the recombiner.
Sample STS 3/4 3-(n+2)
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EH V N E
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, ' Gjneric Letter 89- 01 Enc 1csure 4 H'
TABLE 4.3-9 (Continued)
TABLE NOTATIONS (Not used)
C* During WASTE GAS HOLDUP SYSTEM operation.
.(1) (Not used)
(2) (Not used)
(3) (Not used)
- (4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
- a. One volume percent hydogen, balance nitrogen, and
(5) The CHANNEL CAL.'BRATION shall include the use of standard gas samples containing a nominal:
, s. One volume percent oxygen, balance nitrogen, and
Sample STS 3/4 3-(n+4)
_ __ : _ = _ _ _ _ _ _ _ _ _ - - .__
m- ~. q a
LIST OF RECENTLY ISSUED GENERIC LETTERS
.i Generic. Date of tetter No. Sub.iect Issuance Issued To 38-20 INDIVIDUAL PL ANT 11/23/88 ALL LICENSEES HOLDING EXAMINATION FOR~ SEVERE OPERATING LICENSES ACCIDENT VULNERABILITIES - AND CONSTRUCTION 10 CFR 50.54(f) PERMITS FOR NUCLEAR POWER REACTOR FACILITIES 88-19 USE OF DEADLY FORCE BY 10/28/88 ALL FUEL CYCLE FACILITY LICENSEE GUARDS TO PREVENT LICENSEES WHO POSSESS, THEFT OF SPECIAL NUCLEAR USE, IMPORT. EXPORT, MATERIAL .
OR TRANSPORT FORMULA QUANTITIES OF STRATEGIC SPECIAl NUCLEAR MATERIAL 88-18 PLANT RECORD STORAGE ON 10/20/88 ALL LICENSEES OF OPTICAL DISKS OPERATING REACTORS AND HOLDERS OF CONSTRUCTION PERMITS 88-17 LOSS OF DECAY HEAT REMOVAL 10/17/88 ALL HOLDERS OF 10 CFR S0.54(f) OPERATING LICENSES OR CONSTRUCTION PERMITS FOR PRESSURIZED WATER REACTORS 16 REMOVAL OF CYCLE-SPECIFIC 10/04/88 ALL POWEP REACTOR PARAMETER LIMITS FROM LICENSECS AND TECHNICAL SPECIFICATIONS APPLICANTS 88-15 ELECTRIC POWER SYSTEMS - 09/12/88 ALL POWER REACTOR INADEQUATE CONTROL OVER LICENSEES AND DESIGN PROCESSES APPLICANTS 88-14 INSTRUMENT AIR SUPPLY 08/08/88 ALL HOLDERS OF SYSTEM PROBLEMS AFFECTING OPERATING LICENSES SAFETY-RELATED EQUIPMENT OR CONSTRUCTION PERMITS FOR NUCLEAR POWER REACTORS 88-13 OPERATOR LICENSING 08/08/88 ALL POWER REACTOR EXAMINATIONS LICENSEES AND APFLICANTS FOR AN OPERATING LICENSE.
88-12 REMOVAL OF FIRE PROTECTION 08/02/88 ALL POWER REACTOR REQUIREMENTS FROM TECHNICAL LICENSEES AND SPECIFICATIONS APPLICANTS .
l; ' $ ' '.
- .1 q NUREG-0473
' Revision 3 cy,ECV.ED All ~ '" .
STANDARD RADIOLOGICAL EFFLUENT TECHNICg SPECIFICATIONS FOR BOILING WATER REACTORS 4
e 9
9
+
.--____m._,-._..-_m.__m _m-__._ ___ m__- _.
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A A
- R
- OFFSITE DOSE CALCULATION MANUAL GUIDANCE: j STANDARD RADIOLOGICAL EFFLUENT CONTROLS
- A for BOILING WATER REACTORS
- ' August 1989 A R A t *
- R R A t* A R *
- A A R
- A R R R R A R * * * * * * * * *
- A A A A A R R R A R A ** ** * $t A A A
- R A * * **
- This compilation of Standard Radio 1cgical Effluent Controls (SREC) contains all of the controls required by Generic Letter 89-01, to be incorporated into a licensee's Offsite Dose Calculation Manual (ODCM) at the time the procedural details of the current Radiological Effluent Technical Specifications (RETS) are transferred out of the licensee's Technical Specifi-cations (TS). It has teen developed by recasting the RETS of the most current Standard Technical Specifications from the "LC0" format into the " Controls" format of an ODCM entry. Note that these GE-SREC have been patterned after the W-SREC. The following text guidance incorporates the wording of the most recent SREC; however, no attempt has been made to trans1*te REC numbering of the W-SREC into that of the BWR numbering system.
The following GE-SREC provide the latest version of staff guid- ~
ance, and document current practice in the operating procedures required by 10 CFR 20.106, 40 CFR Part 190, 30 CFR 50.36a and Appendix I to 10 CFR Part 50.
mananman*****************xa*************************************
f L
e j; . .
i TABLE OF CONTENTS h 4, Page-
. '- 1 FOREWORD ............................. ....... .......
1~ DEFINITIONS ............................... ...................
1-0 3/4 . LIMITING CONDITIONS.FOR OPERATION & SURVEILLANCE REQUIREMENTS . 3/4 0-0 3/4.0 A,71icability.................................................. 3/4 0-1 3/4.3 Instrumentation ............................................... 3/4 3-l' 3/4.11 Radioactive Effluents........................................ .. 3/4 11-1 3/4.12 Radiological' Environmental Monitoring .......................... 3/4 12-1
- 3/4' BASES .........................................................
B 3/4 0-0 DESIGN FEATURES ............................................... 5-0
'5
, 6 ADMI NI ST RATIVE CONT RO LS . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
6-0 j: APPENDIX;A': Radiological Assessment' Branch Technical Position Revision'1,. November 1979 APPENDIX B: General Contents of the Offsite Dose Calculation Manual APPENDIX C: Generic Letter 89-01 i l
GE-5 REC Contents t
FOREWORD RADIOLOGICAL EFF'"ENT TECHNICAL SPECIFICATIONS Licensee Technical Specification (TS) amendment requests for incorporation of Radiological Effluent Technical Specifications (RETS) pursuant to 10 CFR 50.36a and Appendix I to 10 CFR Part 50 were approved in the mid-1980s for most operating reactors licensed before 1979 (ors). Plants licensed after 1979 (NTOLs), included the RETS as part of their initial Technical Specifications.
By November 1987, tne RETS were implemented by all licensees of operating power reactors. Detailed Safety Evaluation Reports (SERs) documented the accept-ability of the plant-specific RETS of the ors, while the acceptance of the RETS for the NTOLs followed the regular pattern of the Standard Technical Specifi-cations (STS). Thus, for all operating plants, the compliance of the licensee with 10 CFR 50.36a and Appendix I to 10 CFR Part 50 is a matter of record.
Early draf t revisions of model RETS, distributed to licensees in mid-1978, contained equations for dose calculations, setpoint determinations and meteoro-logical dispersion factors, as well as the procedural details for complying with Appendix I to 10 CFR Part 50. In later revisions, including Revision 2 used as the bench mark for the NRC staff's acceptance of OR RETS, the equations were removed and incorporated into an Offsite Dose Calculation Manual (ODCM) prepared by the licensee and provided to NRC for review along with the proposed RETS.
Early guidance for preparation of the Radiological Effluent Technical Specifi-cations (RETS) and Offsite Dose Calculation Manual (ODCM) was published in NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978. Copies of model RETS, however, have been available only in draft form as NUREG-0472, Revision 2, " Radiological Effluent Technical Specifications for PWRs," February 1,1980; NUREG-0473, Revision 2,
" Radiological Effluent Technical Specifications for BWRs," February 1,1980; and succeeding draft revisions. Staff guidance for the Radiological Environmental Monitoring Program is contained in the Radiological Assessment Branch Technical Position (RAB-BTP), originally issued in March 1978 and upgraded by Revision 1 in November 1979 as a result of the accident at Three Mile Island. This Revision 1 to the RAB-BTP was forwarded to all operating reactor licensees in November 1979 and remains in effect at the present time.
Since this BTP was never incorporated into the Regulatory Guide System, 6 copy is reproduced in this document as Appendix A. Even though it has been used extensively in reviewing ODCMs, guidance for the contents of the ODCM is found only in an appendix to a paper presented at an Atomic Inductrial Forum confer-ence in 1981, and has had only informal distribution since that time.
OFFSITE DOSE CALCULATION MANUAL The potential for augmentation of a licensee's ODCM through transfer of the procedural details of the RETS following the guidance of Generic Letter 89-01, provides an opportunity to assemble in one set of documents the staff guidance for the ODCM.
GE-SREC 1
~
i g ;
The current overview guidance for development of the ODCM was prepared origi-nally in July 1978 and revised in February 1979 after discussions with commit-tees of the Atomic' Industrial Forum. This guidance was made generally available as " Appendix B - General Contents of the Offsite Dose Calculation Manual (ODCM) (Revision 1, February 1979)" to the paper authored by C. A. Willis and F. J. Congel, " Status of NRC P. radiological Effluent Technical Specification Activities" presented at the Atomic Industrial Forum Conference on NEPA and. Nuclear Regulation, October 4-7, 1981, Washington, D.C. A copy of this guidance that continues in effect to date, is reproduced in this document as Appendix B.
During the discussions leading up to the implementation of the RETS by the ors,
- it became important to record in a "living" document certain interpretations and understandings reached in these discussions. The ODCM thus became a repository for such interpretations, as well as for other information requested by the staff in connection with its evaluation of' licensee's commitments and performance under 10 CFR 50.36a and Appendix I to 10 CFR Part 50.
TECHNICAL SPECIFICATION IMPROVEMENT PROGRAM Recently, the NRC staff has examined the contents of the RETS in relation to the Commission's Interim Policy Statement on Technical Specification Improve-ments. The staff has determined that programmatic controls can be implemented in the Administrative Controls section of the Technical Specifications (TS) to satisfy existing regulatory requirements for RETS. At the same time, the procedural details of the current TS on radioactive effluents and radiological environmental monitoring can be relocated to the Offsite Dose Calculation Manual (ODCM). .
To initiate the change, new programmatic controls for radioactive effluents and radiological environmental monitoring are incorporated in the TS to conform to the regulatory requirements of 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part-50. The procedural details included in licensees' present TS on radioactive effluents, environmental monitoring, and
> associated reporting requirements will be relocated to the ODCM. Licensees will handle future changes to these procedural details in the ODCM under the administrative controls for changes to the ODCM. Detailed guidance to effect the transfer of the RETS to the ODCM is given in Generic Letter 89-01, repro-duced in its entirety as Appendix C.
G1H0ANCE FOR_TH,lE TRANSFER OF RETS TO ODCM Erclosure 1 of Generic Letter (GL) 89-01 of Appendix B provides detailed guidance for the preparation of a license amendment request to implement the transfer of RETS to ODCM. Page 1 of the enclosure states:
"The NRC staff's intent in recommending --- the relocation of procedural details of the current RETS to the ODCM is to fulfill the goal of the ,
J p- Commission Policy Statement for Technical Specification Improvements. It is not the staff's intent to reduce the level of radiological effluent control. Rather, this;asendment will provide programmatic controls for RETS consistent with regulatory requirements and allow relocation of the l_ procedaral details of current RETS to the ODCH."
GE-SREC 2
< _ , 'o J EbM ' . _ - _ _
i MlH Page 2 of Enclosure I states:
" ...the procedural details covered in the licensee's current RETS, con-sisting of the limiting conditions for operation, their applicability, remedial actions, surveillance requirements, and the Bases section of the TS for these requirements, are to be relocated to the ODCM --- in a manner that ensures that these details are incorporated in plant operating procedures. The NRC staff does not intend to repeat technical reviews of the relocated procedural details because their consistency with the applicable regulatory requirements is a matter of record from past NRC reviews of RETS."
DISCUSSION For the purpose of the transfer described in GL 89-01 of Appendix B, the RETS will consist of the specifications from the STS listed in Enclosure 2 of Appendix B of GL 89-01. Licensees with nonstandard TS should consider the analogous TS in their format.
It is suggested that the most straightforward method of transferring a licensee's commitments in the RETS to the ODCM in accordance with GL 89-01 is to recast the RETS in the licensee's prescat TS from the " Limiting Condition for Operation (LCO)" format of the TS into the " Controls" format of the ODCM entry. The accompanying package provides an example of this recasting into Standard Radiological Effluent Controls (SREC) from the model RETS for Boiling Water Reactors (BWks). This recasting is in format only. The TS pages have been transferred to the ODCM without change except for the substi- .
tution of " Controls" for "LC0". Plants that have RETS that closely follow the l STS format will be able to use the accompanying examples directly as guidance.
For plants with nonstandard RETS the transfer of TS commitments to the ODCM should be made similarly page by page, again with the substitution of
" Controls" for "LC0".
SUMMARY
As part of the license amendment request for TS improvement relative to the RETS, a licensee confirms that the guidance of Generic Letter 89-01 has been followed. This guidance includes the following:
"The procedural details covered in the licensee's current RETS, consisting of the limiting conditions for operation, their applicability, remedial actions, surveillance requirements, and the Bases section of the TS for these requirements, are to be relocated to the ODCM --- in a manner that ensures that these details are incorporated in plant operating procedures."
The Standard Radiological Effluent Controls (SREC) compiled in this report document current staff practice in the operating procedures required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a and Appendix 1 to 10 CFR Part 50.
{
Thus they contain all of the controls required by Generic Letter 89-01, to be incorporated into a licensee's ODCM at the time the procedural details of the current RETS are transferred out of the licensee's TS.
GE-SREC 3
p
+. 4 2 DMFT 1
NOTE These GE-SREC have been patterned after the W-SREC.
The following text guidance incorporates the wording of the most recent SREC; however, no attempt has been made to translate the REC-numbering of the W-SREC into that of the BWR numbering system GE-SREC 4
9 i %
m M * '
M a
g, i
J SECTION 1.0 DEFINITIONS 4
GE-SREC 1-0
' 1;0 DEFINITIONS The ' defined terms of this 'section appear in capitalized type and are applicable a throughout these Controls.
ACTION 1.1 ACTION shall be that part of a Control that prescribes remedial measures required under designated conditions.
CHANNEL CALIBRATION 1.4 An CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and fhall The CHANNEL CALIBRATION may be_ performed include the CHANNEL FUNCTIONAL TEST.
by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications-and/or status derived from independent instrument channels measuring the same parameter.
CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:
- a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.
- b. Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.
The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.
I
'GE-SREC 1-1
- __ - -__-_ _ _-___ _ _ ______ ___ _ _ _ - _ _ _ _ _ _ __ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - __- _ _ A
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[ DEFINITIONS-
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DOSE EQUIVALENT-I-131 1.10 DOSE EQUIVALENT I-131 shall 'be that concentration of I-131'(microcurie / gram) 3 which alone would produce the same thyroid dose as the quantity and isotopic l
. mixture of I-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid-o -.. dose conversion factors used for this calculation shall be those listed in
-[ Table III of TID-14844, " Calculation of' Distance Factors for Power and Test Reactor Sites" or Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 19773- !
i l
2 GE-SREC 1-2
.l'
i p DEFINITIONS
?'
FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.
GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A SASE005 RADWASTE TREATMENT SYSTEM (e.g., the " augmented offgas system") is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the main condenser evacuation system and providing for delay or holdup for the' purpose of reducing the total radioactivity prior to release to the environment.
MEMBER (S)0FTHEPUBLE 1.16 MEMBER (5) 0F THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.
This category does include persons who use portions of the site for recre-ational, occupational, or other purposes not associated with the plant.
OFFSITE 005E CALCULATION MANUAL 1.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Moni .oring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semiannual Radioactive Effluent Release Reports required by TS 6.9.1.3 and 6.9.1.4.
GE-SREC 1-3
t l l DMFT m
DEQNITIONS OPERABLE - OPERABILITY 1.18 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),
and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water,-lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).
OPERATIONAL CONDITION - CONDITION 1.19 An OPERATIONAL CONDITION, i.e. , CONDITION, shall be any one inclusive combination of. mode switch position and average reactor coolant temperatures as specified in Table 1.2.
I l
PURGE - PURGING ;
1.23 PURGE or PURGING shall bn any controlled process of discharging air or gas !
from a' confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is i required to purify the confinement. l i
l GE-SREC 1-4 .
_ _ _ _ _ __ _ _ _ _ _ _ . _ _ _ _ . . J
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DEFINITIONS c.
. RATED THERMAL' POWER 1.25' RATED TRERMAL P0YER shall be a total reactor core heat transfer rate to
-the reactor coolant of MWt.
REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of.10 CFR Part 50.
l 4
l l
i 1
1 1 '
l SITE BOUNDARY 1.30 The SITE BOUNDARY shall be that line be;<ond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.
GE-SREC 1-5 LL_-_--_____
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.a f T gg
..;;> DEFINITIONS'
.4l-SOURCE CHECK 1.33 A'50URCE CHECK shall be the qualitative assessment of channel response when the channel sensor.is' exposed to a source of increased radioactivity.
%?
THERMAL POWER 1.35 THERMAL POWER shall be the total reactor core heat transfer rate to the reacter coolant.
~
GE-5 REC 1-6 2 - _ _ -__ _ __ - ._. - _ _ . - _ - _
~
~
~
DPET DEFINITIONS UNRESTRICTED AREA 1.38 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.
l VENTILATION EXHAUST TREATMENT SYSTEM 1.39 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particu-lates from the gaseous exhaust stream prior to the release to the environment.
l Such a system is not considered to have any effect on noble gas effluents.
Engineered Safety Features Atmospheric Cleanup.5ystems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.
L VENTING L 1.40 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not pro-vided or required during VENTING. Vent, used in system names, does not imply a VENTING process.
l l
GE-SREC 1-7 1
li W ' .\
ll:
~ TABLE 1.1-1 FREQUENCY NOTATION.
NOTATION FREQUENCY-L 5 At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
L D At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
. W At least once per 7 days.
M At least once per 31 days.
- j. Q At least once per 92 days.
SA At least once per 184 days.
R At least once per 18 months.
1 S/U Prior to each reactor startuo.
N.A. Not applicable.
P Completed prior to each release.
I i
l' GE-SREC 1-8 1
j
l
\ . .
TABLE 1.2 OPERATIONAL CONDITIONS MODE SWITCH AVERAGE REACTOR POSITION COOLANT TEMPERATURE CONDITION Run Any' temperature
- 1. POWER OPERATION Startup/ Hot Standby Any temperature
- 2. STARTL'P Shutdown #' > 200 F
- 3. HOT SHUTDOWN
- 4. COLD SHUTDOWN Shutdown #'##' $ 200*F
$ 140 F
- 5. REFUELING
- Shutdown or Refuel
- 1hc reactor mode switch may be placed in the Run or Startup/ Hot Standby position to test the switch interlock functions provided that the control rods are verified to remain fully inserted by a second licensed operator or other technically qualified member of the unit technical staff.
- The reactor -ode switch may be placed in the Refuel position while a single control rod drive is being removed from the recctor pressure vessel per Specification 3.9.10.1.
- Fuel in the reactor vessel with the vessel head closure bolts I*ss than fully tensioned or with the head removed.
- See Special Tests Exceptions 3.10.1 and 3.10.3.
- The reactor mode switch may be placed in the R'efuel position while a single control rod is being recoupled provided that the one-rod-out interlock is OPERABLE.
GE-SREC 1-9
nm
- l. i SECTIONS 3.0 AND 4.0 CONTROLS AND SURVEILLANCE REQUIREMENTS GE-SREC 3/4 0-0 j
7 q
L 3/4 CONTROLS AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY CONTROLS 3.0.1 Compliance with the Controls contained in the succeeding controls is required during the OPERATIONAL CONDITIONS or other conditions specified therein; except that upon failure to meet the' Control, the associated ACTION requirements shall be met.
3.0.2 Noncompliance with a control shall exist when the requirements of the Control and associated ACTION requirements are not met within the specified time intervals. If the Control is restored prior to expiration of the
' specified time intervals, completion of the ACTION requirements is not required.
3.0.3 When a Control is not met, except as provided in the associated ACTION-requirements, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> action shall be initiated to place the unit in an OPERATIONAL CONDITION in which the control does not apply by placing it, as
! applicable, in:
- 1. At least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- 2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and 1.
- 3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i Where corrective measures are completed that permit operation under the ACTION
. requirements, the action may be taken in accordance with the specified' time limits as measured from the time of failure to meet the Control. Exceptions to these requirements are stated in the individual controls.
This control is not applicable in OPERATIONAL CONDITIONS 4 or 5.
3.0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall not be made unless the conditions for the Control are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through or to OPERATIONAL CONDITIONS as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual controls.
~
1 GE-SREC 3/4 0-1
-- - - __- --m____,,.,__,__m___,___ _
,? -
~
APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1. Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other conditions specified for individual Controls unless otherwise stated in an individual Surveillance Requirement.
4.0.2 Each Surveillance Requirement shall be performed within-the specified time interval with:
- a. A maximum allowable extension not to exceed 25% of the surveillance interval, but
- b. The combined time interval for any three consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance
- interval.
4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a f ailure to meet the OPERABILITY requirements for a Control. Exceptions to these requirements are stated in the individual controls. Surveillance Requirements do not have to be performed on inoperable equipment.
4.0.4 Entry into an OPERATIONAL.IONDITION or other specified applicable condition'shall not be made unless the Surveillance Requirement (s) associated with the Control has been performed within the applicable surveillance interval or as otherwise specified.-
s GE-SREC 3/4 0-2 l
[. ,
[ .
L -
MAFT INSTRUMENTATION i
RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION CONTROLS l
l l
3.3.3.10 In accordance with [ plant name) TS 6.8.4.g.1), the radioactive liquid l
effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Control 3.11.1.1 are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in tbn 0FFSITE DOSE CALCULATION MANUAL (ODCM).
APPLICABILITY _Y: At all times.
ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above control, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.
- b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report pursuant to Control 6.9.1.4 why this inoperability was not corrected in a timely manner.
- c. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
Report all deviations in the Semiannual Radioactive Effluent Release Report.
SURVEILLANCE REQUIREMENTS 4.3.3.10 Each radioactive liquid effluent monitoring instrumentation channel sna11 be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 4.3-8.
GE-SREC 3/4 3-72
N -
O 7 7 8 8 9 I 5 3 3 3 3 3 3 T
C A
SE N MLL O UEB I MNA T I NR 1 1 1 1 1 A NAE 1 T I HP N MCO E
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r m i E e D a a L s l l m e e a I
U Ae e Ae t e n b s n s n t n Q ga i ga e s i i s
I ne u y L L L ne L il l S i
il f s t t E de t de r e n n t V iR n iR f e c e e n I v e v E u u i T of u of ro m t
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n t D f s f t D E n f s g E r E e M so E so e S I
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- R T S
t a in s in r o m s n o s a
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u it v p d e u c i u r e q s t q f t a q t a m t c
cm i cm e o U
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L i
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(NOT USED)
GE-SREC 3/4 3-74
lI A TABLE 3.3-12 (Continued)
ACTION STATEMENTS ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that prior to initiating a l
release:
- a. At least two independent samples are analyzed in accordance
- with Control 4.11.1.1.1, and
- b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving.
Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, grab samples are collected and analyzed for radio-activity at a lower limit of detection of no more than 10 7 microcurie /ml.
ACTION 38 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated l at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump perfor-mance curves generated in place may be used to estimate flow.
ACTION 39 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via -
this pathway may continue provided the radioactivity level is l determined at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. l GE-5 REC 3/4 3-75 i
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DR;FT TABLE 4.3-8 (Continued)
TABLE NOTATIONS (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isola-tion of this pathway and control room alarm annunciation occur if any of the following conditions exists;
- a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or
- b. Circuit failure, or
- c. Instrument indicates a downscale failure, or
- d. Instrument controls not set in operate mode.
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- a. Instrument indicates measured levels above the Alarm Setpoint, or
- b. Circuit failure, or
- c. Instrument indicates a downscale failure, or
- d. Instrument controls not set in operate mode.
(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
(4) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days on which continuous, periodic, or batch releases are made.
I GE-SREC 3/4 3-78
c- - _ - _ - _ _
DR:H INSTRUMENTATION-RJDIDACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION CONTROLS 3.3.3.11 In accordance with [ plant name) TS 6.8.4.g.1), the radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Control 3.11.2.1 are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and -
parameters in the ODCM.
APPLICABILI3: As shown in Table 3.3-13 ACTION:
- a. With a radioactive gaseous effluent monitoring instrumentation channel Alarm / Trip Setpoint less conservative than required by the above control, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.
- b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Semi-annual Radioactive Effluent Release Report pursuant to Control 6.9.1.4 why this inoperability was not corrected in a timely manner.
- c. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
Report all deviations in the Semiannual Radioactive Effluent Release Report.
SURVEILLANCE REQUIREMENTS 4.3.3.11 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies shown in Table 4.3-9.
GE-SREC 3/4 3-79
N O
I 7 1 1 6 6 T 4 5 5 4 4 C
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V y i ea n i ot I S n i n sh o n i o o nx M o dt M T
C n M o r M au A o r M d eE y e R p y D i y e np t l r e n t r e o I
O t
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R i i e S i
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i (NOT USED) l GE-SREC 3/4 3-84
~
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DRUT l
TABLE 3.3-13 (Continued)
TABLE NOTATIONS
- At all times.
- During main condenser offgas treatment system operation.
- During operation of the main condenser air ejector.
ACTION STATEMENTS ACTION 45 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, releases to the environment may continue for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided:
- a. The offgas system is not bypassed, and
- b. The offgas delay system noble gas activity effluent (downstream) monitor is OPERABLE; Otherwise, be in at least HOT STANDBY within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
ACTION 46 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
ACTION 47 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
ACTION 48 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, immediately suspend release of radioactive effluents via this pathway.
ACTION 49 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of the main condenser offgas treatment system may continue provided grab samples are collected at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and analyzed within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
ACTION 50 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, operation of this system may continue for up to 14 days.
ACTION 51 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are contin- !
uously collected with auxiliary sampling equipment as required i in Table 4.11-2. 4 l
l GE-SREC 3/4 3-85 i
4 DRAFT I
l (NOT USED) l GE-SREC 3/4 3-86 I I
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I ct A p e M o rr i
a dis m t l T
C nny s a a e F iP G A aaS a S l t A(
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TABLE 4.3-9 (Contiroed) 1 l
TABLE NOTATIONS l
- At all times.
- During main condenser offgas treatment system operation.
- During operation of the main condenser air ejector.
l (1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarr; annunciation occurs if any of the following conditions exists:
- a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or l
- b. Circuit failure, or
- c. Instrument indicates a downscale failure, or
- d. Instrument controls not set in operate mode.
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:
- a. Instrument indicates measured levels above the Alarm Setpoint, or
- b. Circuit failure, or
- c. Instrument indicates a downscale failure, or
- d. Instrument controls not set in operate mode.
(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related I to the initial calibration shall be used.
(4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
I
(5) The CHANNEL CALIBRATION shall include the use of standard gas samples i j
containing a nominal:
f 7hP ?
lI '.'
3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION CONTROLS 3.11.1.1 In accordance with [ plant name] TS 6.8.4.g.2) and 3), the concentra-tion of radioactive material released in liquid effluents to UNRESTRICTED AREAS '
(see Figure 5.1-3) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concen-tration shall be limited to 2 x 10 4 microcurie /mi total activity.
APPLICABILITY: At all times.
ACTION:
- a. With the concentration of radioactive material released in liquid effluents to UNRESTP.ICTED AREAS exceeding the above limits, immediately restore the concentration to within the above limits.
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
l l
SURVEILLANCE REQUIREMENTS l
4.11.1.1.1 Radioactive liquid wastes shsll be sampled and analyzed according l to the sampling and analysis program of Table 4.11-1. l 4.11.1.1.2 The results of the radioactivity knalyses shall be used in accordance l with the methodology and parameters in the ODCM to assure that the concentrations )
at the point of release are maintained within the limits of Control 3.11.1.1. l l
i GE-SREC 3/4 11-1
l l jl TABLE'4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY (LLD)I )
FREQUENCY ANALYSIS (pCi/ml)-
TYPE FREQUENCY Batch Waste P P l- 1.
Each Batch Each Batch Principal Gamma 5x10 7
! Release Tanks I2) Emitters (3)
I-131 1x10 6 3- n.
P M Dissolved and 1x10 5 One Batch /H Entrained Gases
! (Gamma Emitters) l b. _
P M H-3 1x10 5 Each Batch Composite I4)
Gross Alpha 1x10 7 C:
P Q Sr-89, Sr-90 5x10 8 l
Each Batch Composite (4)
Fe-55 1x10 6 W Principal Gamma 5x10 7
- 2. Continuous Releases (5) Continuous (6) Composite (6) Emitters (3) l 1x10 6 I-131 4.
M M Dissolved and 1x10 5 Grab Sample Entraine1 Gases (Gamma En.itters) b.
M H-3 1x10 5 Continuous (6) Composite (6)
Gross Alpha 1x10 7 C.
Q Sr-89, Sr-90 5x10 8 Continuous (6) Composite (6)
Fe-55 1x10 6 GE-SREC 3/4 11-2
y ("
[.? .!
TABLE NOTATIONS 1)The LLD is-defined,.for purposes of these controls, as the smallest concentration of radioactive material in a sample that will yie_1d a net count, above system background, that will be detected with 95% probability with only-5% probability of . falsely concluding that a blank observation.
reptesents a."real" signal.
L > For a particular measurement system, which may include radiochemical separation:
l 4.66 s b LLD =
E-V 2.22 x 10G Y- exp (-Aat) i Where:
'LLD = the "a priori" lower limit of detection (microcurie per unit mass or volume)',
s = the standard oeviation of the background counting rate or of b
the counting rate of a blank sample as appropriate (counts per minute).
E = the counting efficiency (counts per disintegration),
V = the sample size (units of mass or volume),
2.22 x 106 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclides (sec 1), and at = the elapsed time between the midpoint of sample collection and the time of counting (sec).
Typical values of E, V, Y, and at should be used in the calculation.
It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.
I2)A batch release is the discharge of liquid wastes of a discrete volume.
Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed by a method described in the ODCM to assure representative sampling.
GE-SREC 3/4 11-3
b
' , ji c.
b TABLE 4.11-1 (Continued)
TABLE NOTATIONS (Continued)
(3)The principal gamma emmiters for which the LLD control applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134, Cs-137, and Ce-141. Ce-144 shall also he measured, but with an LLD of 5 x 10 6 This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Control 6.9.1.4 in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974.
( )A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.
(5)A continuous release is the discharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release.
(6)To be representative of the quantities and concentrations of radioactive ,
materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.
GE-SREC 3/4 11-4 l
J '
DR H RADI0 ACTIVE EFFLUENTS DOSE CONTROLS 3.11.1.2 In accordance with [ plant name) TS 6.8.4.g.4) and 6.8.4.9.5), the dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Figure 5.1-3) shall be limited:
- a. During any calendar quarter to less than or equal to 1.5 mrems. to the whole body and to less than or equal to 5 mrems to any organ, and
- b. During any calendar year to less than or equal to 3 mrems to the whole body and to'less than or equal to 10 mrems to any organ.
APPLICABILITY: At all times.
ACTION:
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Control 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been 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. This Special Report shall also include: (1) the results of radiological analyses of the drinking water source, and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR Part 141, Safe Drinking Water Act.*
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS .
4.11.1.2 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least,once per 31 days.
- The requirements of ACTION a.(1) and (2) are applicable only if drinking water supply is taken from tne receiving water body within 3 miles of the plant discharge. In the case of river-sited plants this is 3 miles downstream only.
GE-SREC 3/4 11-5
t DRJT RADIOACTIVE EFFLUENTS LIQUID RADWASTE TREATMENT SYSTEM CONTROLS
.3.11.1.3 In accordence with [ plant name) TS 6.8.4.g.6), the Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the system shall
' e used to reduce releases of radioactivity when the projected doses due to the o
liquid effluent, from each unit, to UNRESTRICTED AREAS (see Figure 5.1-3) would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31-day period.
APPLICABILITY: At all times.
ACTION:
- a. With radioactive liquid waste being ditcharged without treatment and in excess of the above limits and any portion of the Liquid Radwaste Treatment System not in operation, prepare and submit to the Commis-sion within 30 days, pursuant to Control 6.9.2, a Special Report that includes the following information:
- 1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence,
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases from each unit to UNRESTRICTED AREAS -
shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Liquid Radwaste Treatment Systems are not being fully utilized.
4.11.1.3.2 The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting Controls 3.11.1.1 and 3.11.1.2.
GE-SREC 3/4 11-6
W ,
9 p
3/4.11.1.4 (NOT USED) p 9
GE-SREC 3/4 11-7 u---___
RADI0 ACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS DOSE RATE CONTROLS 3.11.2.1 In accordance with [ plant name] TS 6.8.4.g.3) and 7), the. dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following:
- a. For noble gases: Less than or equal to 500 mrems/yr to the whole body and less than or equal to 3000 mrems/yr to the skin, and
- b. For Iodine-131, for Iodine-133, for tritium, and for all radio-nuclides in particulate form with half-lives greater than 8 days:
Less than or equal to 1500 mrems/yr to any organ.
APPLICABILITY: At all times.
ACTION:
- a. With the dose rate (s) exceeding the above limits, immediately restore the release rate to within the above limit (s).
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.1.1 The dose rate due to noble gases in pseous effluents shall be determined to be within the above limits in accordance with the snethodology and parameters in the ODCM.
4.11.2.1.2 The dose rate due to Iodine-131 Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 4.11-2.
-GE-SREC 3/4 11-8
I I
)
T(l 1
~
m " 1 1
~
I MN/ 4 4 8 4 1 1 1 8
- - - - 8 - - - - -
I O1 0 0 0 0 LIC 0 0 0 0 0 T p 1 1 1 1 01 1 1 1 1 RC( x x x x 1 x x x x x EE 1 1 1 1 x1 1 1 1 1 WT 1 -
OE -
LD
) ) ) )
M 2 2 2 2 A ( ( ( (
R s s s s G r r r r O e e e e R t t t t P t t t t i i i i a S m m m m m I S E E E E m S I a Y S a a a a G L Y m m m m r A L m m m m o N FA a a a a 0 A ON G G ) G ) G a 9 s A e e h - ea D E l l d l d l p F st N PY a a i a i a l S ae A YT p p x p x p A GB TI i i o i o i ,
es G V c c ( c (1 c s 9 2 N I n n n 3 n s 8 l s
- I T i i 3 i 31 i o - bo 1 L C r r - r - - r r r or 1 P A P P H P Hl P G S NG
. M e e 4 A l - l S ) rp rp E 3 am am L E I e P a P a B T E t S S A S Y G ) ' ) a e e s T A SC R 3 n' l 7 l t e t e a W MIN U (
,'W a ( u it it Gr USE M PP M M o e yce Msa Qsa e S MYU cl il ol ol et U I LQ h rp t p pu pu l i O NAE c am rm mc mc b n oo E I NR a h a aa oi oi S MAF E CS PS Ct Ct NM A
G ) ) ) ) )
, 6 6 6 6 E 3 ," (
V e ( e e s
(
s
(
s
(
s s I l El l T Y p Gp p u u u u u C GC m Rm m o o o o o A NN a Ua a u u u u u 0 I E S PS S n n n n n i i 1 LU M P i i i t t 0 PQ b hb ) b t t t A ME a ca 3 a n n n n n R AR r ar I r o o o o o SF G EG M G C C C C C s
t sem)
E E nuroy s P t G ioart e .e Y n R roe fi p .v T e U epss l y1 o m P h atdi T b E t t egnec na S a t os esa ei A e n aeuaf s .
E r e t erl e ad3 L T m sl efl e ee E nT i ehf eh l td R sm iN L rwert esn ae aE ( Ri a S gt tV l U f s n l l s O f y oR . .
E OS CO a Aa2 S
A . . .
G 1 2 3 .
4 onrJ ,M" wD gI* '
TABLE 4.11-2 (Continued)
TABLE NOTATIONS (1)The LLD is defined, for purposes of these controls, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system, which may include radiochemical separation:
4" $
D LLD =
E V 2.22 x 106 Y exp (-Aat)
Where:
LLD = the "a priori" lower limit of detection (microcurie per unit mass or volume),
s = the standard deviation of the background counting rate or of b
the counting rate of a blank sample as appropriate (counts per minute),
E = the counting efficiency (counts per disintegration),
V = the sample size (units of mass or volume),
2.22 x 106 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclides (sec 1), and i at = the elapsed time between the midpoint of sample collection and the time of counting (sec).
Typical values of E, V, Y, and at should be used in the calculation.
l It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (af ter the f act) limit for a particular measurement.
I GE-SREC 3/4 11-10
ggg73 TABLE 4.11-2 (Continued) f TABLE NOTATIONS (Continued)
(2)The principal gamma. emitters for which the_LLD control applies include the following radionuclides: Kr-87, Kr-88, Xe-133,' Xe-133m, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, In-65, Mo-99, 1-131,-Cs-134, Cs-137, Ce-141 and Ce-144 in Iodine and particulate releases..This' list does not mean that only these rtclides.are to be
. considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in.the-Semiannual Radioactive Effluent Release Report pursuant to Control 6.9.1.4
.in the format outlined in Regulatory Guide 1.21,' Appendix B, Revision 1, June 1974.
I3) Samplingand analysis shall also be performed following shutdown .startup, or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period.
(4)Not applicable.
(5) Tritium grab samples shall be taken at'least once per 7 days from the ventilation exhaust from the_ spent fuel pool area, whenever spent fuel is in the spent fuel pool.
IO)The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Controls 3.11.2.1, 3.11.2.2, and 3.11.2.3.
I ) Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing, or.after removal from sampler.
Sampling shall also be performed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for at least 7 days following each shutdown, startup, or THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a 1-hour period and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of changing. When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are analyzed, the corresponding LLDs pay be increased by a factor of 10.
This requirement does not apply if: (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the reactor coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not-increased more than a factor of 3.
1 GE-SREC 3/4 11-11
RADI0 ACTIVE EFFLUENTS DOSE - NOBLE GASES CONTROLS 3.11.2.2 In accordance with [ plant name] TS 6.8.4.g.5) and 8), the air dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following:
- a. During any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radiation, and
- b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less than or equal to 20 mrads for beta radiation.
APPLICABILITY: At all times.
ACTION
- a. With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Control 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been 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.
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.2 Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. _
GE-SREC 3/4 11-12
DMFT
, RADIOACTIVE EFFLUENTS I
DOSE - 10 DINE-131,10 DINE-133, TRITIUM, AND RADI0 ACTIVE MATERI AL IN l PARTICULATE FORM CONTROLS 3.11.2.3 In accordance with [ plant name) TS 6.8.4.g.5) and 9), the dose to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radio-ruclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) shall be limited to the following:
- n. During any calendar quarter: Less than or equal to 7.5 mrems to any organ and,
.b. During any calendar year: Less than or equal to 15 mrems to any organ.
APPLICABILITY: At all times.
ACTION:
- a. With the calculated dose from the release of Iodine-131, Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Control 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been 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.
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 11.2.3 Cumulative dose contributions for the current calendar quarter and current calendar year for Iodine-131, Iodine-133, tritium and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.
GE-SREC 3/4 11-13
DRAFT RADI0 ACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT SYSTEM CONTROLS
'3.11.2.4 The GASEOUS RADWASTE TREATMENT SYS'EM shall be in operation.
APPLICABILITY: - Whenever the main condenser air ejector (evacuation) system is
,in operation.
^ ACTION:
- a. With gaseous radwaste from the main condenser air ejector system being discharged without treatment for more than 7 days, prepare and submit to the Commission within 30 days, pursuant to Control 6.9.2, a Special Report that includes the following information:
- 1. Identification of the inoperable equipment or subsystems and the reason for inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary, description of action (s) taken to prevent a recurrence.
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.4 The readings of the relevant instruments shall be checked every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the main condenser air ejector is in use to ensure that the gaseous radwaste treatment system is functioning.
.1 I
GE-SREC 3/4 11-14
DRE RADI0 ACTIVE EFFLUENTS VENTILATION EXHAUST TREATMENT SYSTEM CONTROLS 3.11.2.5 The VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE and appropriate portions of this system shall be used to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 5.1-3) would exceed:
- a. 0.2 mrad to air from gamma radiation, or
- b. 0.4 mrad to air from beta radiation, or
- c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.
APPLICABILITY: At all times.
ACTION:
- a. With radioactive gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Control 6.9.2, a Special Report that includes the following information:
- 1. Identification of any inoperable equipment or subsystems, and the reason for the inoperability,
- 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and
- 3. Summary description of action (s) taken to prevent a recurrence,
- b. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
l SURVEILLANCE REQUIREMENTS 4.11.2.5.1 Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when the Ventilation Exhaust Treatment System is not being fully utilized. ,
I 4.11.2.5.2 The installed VENTILATION EXHAUST TREATMENT SYSTEM shall be considered OPERABLE by meeting Controls 3.11.2.1, and either 3.11.2.2 or 3.11.2.3.
i l
GE-SREC 3/4 11-15 l l
u- - _ _ - - - - _ - - - - - - - - - - - - - - - - - . J
, = . _ _ _ _ ,. _ _ _ _ . . _ _ . _ _ . . . .
DRLFI RADIOACTIVE EFFLUENTS 3/4.11.2.6 (NOT USED)
I GE-5 REC 3/4 11-16 U __-
DMFi RADI0 ACTIVE EFFLUENTS 3/4.11.2.7 (NOT USED)
)
i GE-SREC 3/4 11-17
DRAFT RADIOACTIVE EFFLUENTS l ..
MARK I or II CONTAINMENT-CONTROLS 3.11.2.8 VENTING or PURGING of the' Mark I or 11 containment drywell shall be
-through the Standby _ Gas. Treatment System.
' APPLICABILITY: Whenever the drywell is vented or purged.
' ACTION:
- a. With'the requirements of the above control not satisfied, suspend all VENTING and PURGING of the drywell.
- b. Theprovisionsoffontrols3.0.3and3.0.4arenotapplicable.
SURVEILLANCE REQUIREMENTS 4.11.2.8 The containment drywell shall be determined to be aligned for VENTING or PURGING through the Standby Gas Treatment System within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to start of and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during VENTING or PURGING of the drywell.
l
{
l GE-SREC 3/4 11-18
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l l RADIOACTIVE EFFLUENTS 3/4.11.3 (NOT USED)
GE-SREC 3/4 11-19 L__.__.__ . . . . _ _ _ _ _ _ O
l RADI0 ACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE CONTROLS 3.11.4 In accordance with [ plant name] TS 6.8.4.g.11), the annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems.
APPLICABILITY: At all times.
ACTION:
- a. With the calculated doses from the release of radioactive caterials in liquid or gaseous effluents exceeding twice the limits of Control 3.11.1.2a., 3.11.1.2b., 3.11.2.2a. 3.11.2.2b., 3.11.2.3a., or 3.11.2.3b., calculations shall be made including direct radiation contributions from the units (including outside storage tanks etc.) to determine whether the above limits of Control 3.11.4 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Control 6.9.2, a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits.
This Special Report, as defined in 10 CFR 20.405(c), shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive raterial involved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the above limits, and if the release condition result-ing in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accor-dance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete.
- b. The provisions of Contruls 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Controls 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the methodology and parameters in the ODCM.
4.11.4.2 Cumulative dose contributions from direct radiation from the units (including outside storage tanks etc.) shall be determined in accordance with the methodology and parameters in the ODCM. This requirement is applicable cnly under conditions set forth in ACTION a. of Control 3.11.4.
GE-SREC 3/4 11-20
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3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1- MONITORING PROGRAM CONTROLS-3.12.1 In accordance with [ plant name) TS 6.8.4.h.1), the Radiological Environmental Monitoring Program shall-be conducted as specified in Table 3.'12-1.
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APPLICABILITY: At all times.
ACTION:
- a. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 3.12-1, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Control 6.9.1.3, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.
- b. With the level of radioactivity as the result of plant effluents in an environmental' sampling medium at a specified location exceeding the reporting levels of. Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days, pursuant to Control 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be
'taken to reduce radioactive effluents so that the potential annual dose
- to a MEMBER OF THE PUBLIC is less than the calendar year limits of Controls 3.11.1.2, 3.11.2.2, or 3.11.2.3. When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:
concentration (1) concentration (2) + ***> 1.0 reporting level (1) , reporting level (2) -
When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose
- to a MEMBER OF THE PUBLIC from all radio-nuclides is equal to or greater than the calendar year limits of Control 3.11.1.2, 3.11.2.2, or 3.11.2.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report required by Control 6.9.1.3.
- The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.
CE-SREC 3/4 12-1 l
~
RADIOLOGICAL ENVIRONMENTAL MONITORING CONTROLS ACTION (Continued)
- c. With milk or fresh leafy vegetation samples unavailable from one or more of the sample locations required by Table 3.12-1, identify specific locations for obtaining replacement samples and add them within 30 days to the Radiological Environmental Monitoring Program given in the ODCM. The specific locations from which samples were unavailable may then be deleted from the snonitoring program. Pursuant
~
to Control 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table for the ODCM reflecting the new location (s) with supporting information identifying the cause of the unavailability of samples and justifying the selection of the new location (s) for obtaining samples.
- d. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
_ SURVEILLANCE REQUIREMENTS 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table 3.12-1 from the specific locations given in the table and figure (s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 3.12-1 and the detection capabilities required by Table 4.12-1.
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'lABLE 3.12-1 (Continued)
TABLE NOTATIONS (1) Specific parameters of distance and direction sector from the centerline of one reactor, and additional description where pertinent, shall be pro-vided for each and every sample location in Table 3.12-1 in a table and figure (s) in the ODCM. Refer to NUREG-0133, " Preparation of Radiological Ef fluent Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Technical Position, Revision 1, November 1979. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to circumstances such as hazardous conditions, seasonal unavailability, and malfunction of auto-matic sampling equipment. If specimens are unobtainable due to sampling equipment malfunction, effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environ-mental Operating Report pursuant to Control 6.9.1.3. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program given in the ODCM. Pursuant to Control 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table for the ODCM reflect-ing the new location (s) with supporting information identifying the csuse of the unavailability of samples for the pathway and justifying the selec-tion of the new location (s) for obtaining samples.
(2) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addi-
! tion to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters.
Film badges shall not be used as dosimeters for measuring direct radiation.
(The 40 stations is not an absolute number. The number of direct radiation monitoring stations may be reduced according to geographical limitations; e.g., at an ocean site, some sectors will be over water so that the number of dosimeters may be reduced accordingly. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the speci-fic system used and should be selected to obtain optimum dose information with minimal fading.)
(3) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or more af ter sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.
l GE-SREC 3/4 12-7 l
L__ _. _ . . _ _ _ _ _ _ _ _ _ _ _ _ . _ _
l l A
-TABLE 3.12-1 (Continued)
TABLE NOTATIONS (Continued)
(4) Gamma' isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.
(5) The " upstream sample" shall be taken at a distance beyond significant influence of the discharge. The " downstream" sample shall be taken in an area beyond but near the mixing zone. " Upstream" samplis in an estuary must be taken far enough upstream to be beyond the plant influence. Salt water shall be sampled only when the receiving water is utilized for. ,
recreational activities.
(6) A composite sample is one in which the quantity (aliquot) of liquid sampled is proportional to the quantity of flowing liquid and in which the method of sampling employed results in a specimen that is representative of the liquid flow. In this program composite sample aliquots shall be collected at time intervals that are very short (e.g. , hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample.
(7) Groundwater sainples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination.
(8) The dose shall be calculated for the n.sximum organ and age group, using the methodology and parameters in the ODCM.
(9) If harvest occurs more than once a year, sampling shall be performed during each discrete harvest. If harvest occurs continuously, sampling shall be scnthly. Attention shall be paid to including samples of tuberous and root food products.
9 GE-SREC 3/4 12-8
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< b TABLE 4.12-1 (Continued)
TABLE NOTATIONS (1)This list does not mean that only these nuclides are to be considered.
Other peaks that are identifiable, together with those of the above .
nuclides, shall also be analyzed and reported in-the Annual Radiological Environmental Operating Report pursuant to Control 6.9.1.3.
(2) Required detection capabilities for thermoluminescent dosimeters used.
for environmental measurements shall be in accordance with the recommenda-tions of Regulatory Guide 4.13.
(3)The LLD is defined, for purposes of these controls, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system, which may include radiochemical separation:
- 4. 6 s b
LLD =
E - V -
2.22 - Y - exp(-AAt)
Where:
LLD = the "a priori" lower limit of detection (picoCuries per unit mass or volume), ,
s b
= the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute),
E = 'he counting efficiency (counts per disintegration),
V = the sample size (units of mass or volume),
2.22 = the number of disintegrations per minute per picocurie, Y = the fiactional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclides (sec 2), and At = the elapsed time between environmental collection, or end of the sample collection period, and time of counting (sec).
Typical values of E, V, Y, and At should be used in the calculation.
GE-SREC 3/4 12-11
- DRAff TABLE 4.12-1 (Continued)
, TABLE NOTATIONS (Continued)
It should be. recognized that the LLD is defined as~an a priori (before the.
fact) limit representing the capability of a measuremelit systens and not as an a_ posteriori (after the fact) limit for a particular measurement.
Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable.
In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Control 6.9.1.3.
i 1
=
GE-SREC 3/4 12-12 L_- _ _ _ _ _ _ _ _ _ _ _ _ .
L RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS CONTROLS 3.12.2 In accordance with [ plant name] TS 6.8.4.h.2), a Land Use Census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence, and the nearest garden
- of greater than 50 2m (500 ft2 )
producing broad leaf vegetation. [For elevated releases as defined in Regula-tory Guide 1.111, Revision 1, July 1977, the Land Use Census shall also identify within a distance of 5 km (3 miles) the locations in each of the 16 meteorological sectors of all milk animals and all gardens of greater than 50 m2 producing broad leaf vegetation.]
APPLIC/BILITY: At all times.
ACTION:
- a. With a Land Use Census identifying a location (s) that yields a calculated dose or dore commitment greater than the values currently being calculated in Control 4.11.2.3, pursuant to Control 6.9.1.4, identii'y the new location (s) in the next Semiannual Radioactive Effluent Release Report.
- b. With a Land Use Census identifying 3 location (s) that yields a calculated dose or dose comt'tment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with Control 3.12.1, add the new location (s) within.30 days to the Radiological Environmental Moni-toring Program given in the ODCM. The sampling location (s), exclud-ing the control station location, having the lowest calculated dose or dose commitment (s), via the same exposure pathway, may be deleted from this monitoring program after [0ctober 31] of the year in which this Land Use Census was conducted. Pursuant to Control 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table (s) for the ODCM t, flecting the new location (s) with informa-tion supporting the change in sampling locations.
- c. The provisions of Controls 3.0.3 and 3.0.4 are not applicable.
- Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census.
Controls for broad leaf vegetation sampling in Table 3.12-1, Part 4.c., shall be followed, including analysis of control samples.
GE-SREC 3/4 12-13
--.__._.____.-______m_____ . _ _ _ - _ _ _ _ _ _ _ _ _ _ . _ _ - _ _ _ _ _ - . . . _ . _ . . _ _ _
b
' RADIOLOGICAL ENVIRONMENTAL MONITORING -
SURVEILLANCE REQUIREMENTS 4.12.2 The Land Use Census shall be conducted during the growing season at least once per 12 months ucing that information that will provide.the best results, such as by a door-to-door survey, aerial survey, or by consulting
- local agriculture authorities. The results of the Land Use Census shall be
' included in the Annual Radiological Environmental Operating Report pursuant to Control 6.9.1.3.
P I
w 4
i GE-5 REC 3/4 12-14
~
2 e DRAFT RT RADIOLOGICAL ENVIRONMENTAL MONITORING
, 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM l CONTROLS ~
3.12.3 In accordance with [ plant name).TS 6.8.4.h.3),-analyses shall be
. performed on all radioactive materials, supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission, that-correspond to
- samples required by Table 3.12-1.
APPLICABILITY: 'At all times.
ACTION:.
t-
- a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the. Commission in the Annual' Radiological Environmental Operating Report pursuant
-to Control 6.9.1.3.
b .- The provisions of Controls 13.0 .3 and 3.0.4 are not applicable.
SURVEILLANCE REQ'JIREMENTS 4.12.3 The Interlaboratory Comparison Program shall be described in the ODCM.
A summary of the results obtained as part of'the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Control 6.9.1.3.
GE-SREC 3/4 12-15
_m _..- -m___._m.__-__.__.______ . _ _ - . . _ . .
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DRlFT I
BASES FOR SECTIONS 3.0 AND 4.0 CONTROLS AND l SURVEILLANCE REQUIREMENTS NOTE The BASES contained in succeeding pages summarize the reasons for the Controls in Sections 3.0 and 4.0, but are not part of these Controls.
l l
I
- . /
INSTRUMENTATION I
BASES "3/4.3.3.10 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements -
of General- Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
3/4.3.3.11 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous efflu-ents during actual or potential releases of gaseous effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirem2nts of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
GE-SREC B 3/4 3-6
l l
- : DRET l
3/4.11 RADI0 ACTIVE EFFLUENTS BASES 3/4.11.1 LIQUID EFFLUENTS 3/4.11.1.1 CONCENTRATION This control is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II, Column 2. This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within: (1) the Section II.A design objectives of 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. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission en Radiological Protection (ICRP) Publication 2.
This control applies to the release of radioactive materials in liquid effluents from all units at the site.
The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs).
Detailed discussion of the LLD, and other detection limits can be found in Currie, L. A. , " Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements,"
NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300.
3/4.11.1.2 DOSE This control is provided to implement the requirements of Sections II.A, III.A, and IV.A of Appendix I, 10 CFR Part 50. The Control implements the guides set forth in Section II.A of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix I to assure that the releases of radio-active material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." Also, for fresh water sites with drinking water supplies that can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclides concentrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. The dose calculation methodology and parameters in the ODCM implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calcu-lational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substan-tially underestimated. The equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid effluents are consistent with the methodology provided in Regulatory Guide 1.109. " Calculation of Annual Doses to Man from Routine Releases of GE-SREC B 3/4 11-1
- : DWFT l l
i RADIOACTIVE EFFLUENTS BASES DOSE (Continued) ,
Reactor Ef fluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix 1," Revision 1, October 1977 and Regulatory Guide 1.113. " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.
This control applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared Radwaste Systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system.
3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in liquid effluents will be kept "as low as is reasonably achievable." This control implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the Liquid Radwaste Treatment System were specified as a suitable fraction of the dose design objectives set forth in Section II. A of Appendix 1,10 CFR Part 50 for liquid effluents.
This control applies to the release of radioactive materials in liquid effluents from each unit at the site. For units with shared Radwaste Systems, the liquid effluents from the shared system are to be proportioned among the-units sharing that system.
GE-SREC B 3/4 11-2
i l l i
RADI0 ACTIVE EFFLUENTS BASES l
3/4.11.2 GASEOUS EFFLUENTS j 3/4.11.2.1 DOSE RATE This control is provided to ensure that the dose at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 to UNRESTRICTED AREAS.
The annual oose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1. These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 For MEMBERS OF THE PUBLIC who may at times be within (10 CFR Part 20.106(b)).
the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that ior the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCH. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrems/ year to the whole body or to less than or equal to 3000 mrems/ year to the skin.
These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 mrems/ year.
This control applies to the release of radioactive materials in gaseous effluents from all units st the site.
The required detection capabilities for radioactive material in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs).
Detailed discussion of the LLD, and other detection limits can be found in Currie, L. A. , " Lower Limit of Detection: Definition and Elaboration of a Proposed Position for Radiological Effluent and Environmental Measurements,"
HUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300.
3/4.11.2.2 DOSE - NDBLE GASES This control is provided to implement the requirements of Sections II.B, III. A and IV. A of Appendix 1,10 CFR Part 50. The control implements the guides set forth in Section I.B of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix I to assure that the releases of radio-active material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through' appropriate pathways is unlikely to be substantially underestimated. The dose calculation GE-SREC B 3/4 11-3
1 i
DAFT i i
RADI0 ACTIVE EFFLUENTS BASES i
j DOSE-NOBLE GASES (Continued) methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, <
" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents l for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I,"
Revision I, October 1977 and Regulatory Guide 1.111. " Methods for Estimating Atmospheric Transport And Dispersion of Gaseous Effluents in Routine Releases f rom Light-Water CW kd Reactors," Revision 1, July 1977. The ODCM equations provided for deterA sng the air doses at and beyond the SITE BOUNDARY are based upon the histo, kal average atmospheric conditions.
This control applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.
3/4.11.2.3 DOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADIOACTIVE MATERIAL IN PARTICULATE FORM This control is provided to implement the requirements of Sections II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The Controls are the guides set forth in Section II.C of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV. A of Appendix I to assure that the releases of radioactive materials in gcseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calcu-lational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substan-tially underestimated. The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 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,"
Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1. July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate controls for Iodine-131 Iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days are dependent upon the existing radionuclides pathways to man inthe GE-SREC B 3/4 11-4 L-_---_-----_____
DMFT
, RADIOACTIVE EFFLUENTS BASES DOSE - 10 DINE-131, 10 DINE-133, TRITIUM, AND RADIOACTIVE MATERIAL IN PARTICULATE p
FORM (Continued) areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of the calculations were: (1) individual inhalation of airborne radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat
- by man, and (4) deposition on the ground with subsequent exposure of man.
This control applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.
3/4.11.2.4 AND 3/4.11.2.5 GASEOUS RADWASTE TREATMENT SYSTEM AND VENTILATION EXHAUST TREATMENT SYSTEM The OPERABILITY of the WASTE GAS HOLDUP SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This control implements the requirements of 10 CFR 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.
This control applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared Radwaste Treatment-Systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.
I i
4 GE-5 REC B 3/4 11-5 l
1
- _ _ _ _ _ _ _ - _ . _ _ _ _ _ _ _ _ 1
7.,,._
[ Q RADIOACTIVE EFFLUENTS-
~
BASES
-3/4 11.2.6 NOT USED 3/4 11.2.7' NOT USED 3/4.11.2.8 MARK I CONTAINMENT This specification provides reasonable assurance that releases from drywell
_ purging operations will not exceed the annual dose limits of 10 CFR part 20 for unrestricted areas.
3/4.11.3 NOT USED 3/4.11.4 TOTAL DOSE This :ontrol is provided to meet the dose limitations of 10 CFR Part 190
- that have neen incorporated into 10 CFR Part 20 by 46 FR 18525. The control requires the preparation and submittal of a Special Report whenever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ,
' except the thyroid, which shall be limited to less than or equal to 75 nrems.
For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I and if direct radiation doses from the units GE-SREC B 3/4 11-6
DRAFT RADI0 ACTIVE EFFLUENTS BASES
-TOTAL DOSE (Continued)
(including outside storage tanks, etc.) are kept small. The Special Report will describe a course of action that should result in the limita?. ion of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits.
For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER of the PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at'the same site or within a radius of 8 km must be
' considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in-violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is com-pleted. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in Controls 3.11.1.1 and 3.11.2.1. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.
r GE-SREC B 3/4 11-7
DRAFT 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM f The Radiological Environmental Monitoring Program required by this cuntrol provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposure of MEMBERS OF THE PUBLIC resulting from the plant operation. This monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby 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 the environmental {
exposure pathways. Guidance for this monitoring program is provided by the j Radiological Assessment Branch Technical Position on Environmental Monitoring, l Revision 1, November 1979. The initially specified monitoring program will be j effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience.
The required detection capabilities for environmental sample analyses are tabulated in terms of the lower limits of detection (LLDs). The LLDs required by Table 4.12-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measure-ment system and not as an a posteriori (after the fact) limit for a particular measurement.
I Detailed discussion of the LLD, and other detection limits, can be found i in Currie, L. A., " Lower Limit of Detection: Definition and Elaboration of a !
Proposed Position for Radiological Effluent and Environmental Measurements,"
NUREG/CR-4007 (September 1984), and in the HASL Procedures Manual, HASL-300.
3/4.12.2 LAND USE CENSUS This control is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program given in the ODCM are made if required by the results of this census. The best information from the door-to-door survey, from aerial survey or from consulting with local agri-cultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 50 m2 provides assurance that significant exposure i pathways via leafy vegetables will be identified and monitored since a garden l of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child.
To determine this minimum garden size, the following assumptions were made: ;
(1) 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m2.
GE-SREC B 3/4 12-1 l
~; .
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. . RADIOLOGICAL ENVIRONMENTAL ~ MONITORING i -: . BASES
-3/4.12.3 INTE'RLABORATORY COMPARISON PROGRAM' The requirement for participation in'an approved Interlaboratory Comparison Program is provided to ensure that independent' checks on the precision and accuracy of the measurements of radioactive materials in' environmental sample matrices are performed as part of the quality assurance. program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix I to 10 CFR Part 50. 9 G
S mnemo O
GE-SREC B 3/4 12-2
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, iO, I k.I i-t SECTION 5.0 DESIGN FEATURES !
GE-SREC 5-0 !
l l
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. 5.0 DESIGN FEATURES 1 5.1~ SITE l
MAP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADI0 ACTIVE GASEOUS AND MQUIDEFFLUENTS 5.1.3 Information regarding radioactive gaseous and liquid effluents, which will allow identification of structures and release points as well as defini-tion of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC, shall be as shown .in Figure [5.1-3). -
The definition of UNRESTRICTED AREA used in implementing these Controls has been expanded over that in 10 CFR 20.3(a)(17) The UNRESTRICTED AREA boundary may coincide with the Exclusion (fenced) Area boundary, as defined in 10 CFR 100.3(a), but the UNRESTRICTED AREA does not include areas over water-bodies. The concept of UNRESTRICTED AREAS, established at or beyond the SITE BOUNDARY, is utilized in the Controls to keep levels of radioactive materials in liquid and gaseous effluents as ^.ow as is reasonably achievable, pursuant to 10 CFR 50.36a.
c GE-SREC 5-1
..' I-i.
This figure shall consist _of a map of.the site area-l ' showing the SITE BOUNDARY and locating points within i the SITE BOUNDARY where radioactive gaseous and liquid
' effluents are' released, as well as where radioactive liquid effluents 10 ave the site. If onsite areas sub-ject to radioactive materials in gaseous or liquid - .
effluents are utilized by the public for recreational or other purposes, these areas shall be outlined on the map and identified by occupancy control (if any).
The figure shall be sufficiently detailed to allow identification of structures and release point locations and elevations, as well as definition of UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC. The map scale shall be on the order of 2-3"/ mile. See NUREG-0133 for additional guidance.
FIGURE 5.1-3 UNRESTRICTED AREA AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS GE-SREC 5-4
. , . - - . - - - - , , , - - , , . - . -- - - - . - - - , - - - . - - - - - , - - - - - - - -- - - . - - -----.-.,r--. . --
(jf ::.- .'f f , ADMINISTRATIVE CONTROLS i
o ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT **
6.9.1.3 Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each. year. The initial report shall be submitted prior to May 1 of the year following initial criticality and shall include copies of reports of the preoperational Radiological Environmental Monitoring Program of the' unit for at least'two years prior to initial criticality.
L The Annual-Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the i radiological environmental. surveillance activities for the report period,
- A single submittal may be made for a multiple unit station.
GE-SREC 6-17
3 Riff ADMINISTRATIVE CONTROLS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT (Continued) including a comparison with preoperational studies and with operational controls, as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of the Land Use Census required by Control 3.12.2.
The Annual Radiological Environmental Operating Reports shall include the i results of analysis of all radiological environmental samples and of all
) environmental radiation measurements taken during the period pursuant to the l locations specified in the table and figures in the Offsite Dose Calculation l
Manual, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some indivi-dual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report.
The reports shall also include the following: a summary description of the Radiological Environmental Monitoring Program; at least two legible maps
- covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program and the corrective action taken if the specified program is not being performed as required by Control 3.12.3; reasons for not conducting the Radiological Environmental Monitoring Program as required by centrol 3.12.1, and discussion of all deviations from the sampling schedule of Table 3.12-3; discussion of environmental tample measure-ments that exceed the reporting levels of Table 3.12-2 but are not the result of plant effluents, pursuant to ACTION b. of Control 3.12.1; and discussion of all analyses in which the LLD required by Table 4.12-1 was not achievable.
SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT **
6.9.1.4 Routine Semiannual Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period --
of the first report shall begin with the date of initial criticality.
The Semiannual 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," Revision 1 June 1974, with data
- 0ne map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations. l I
- A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal tha11 l specify the releases of radioactive material from each unit.
GE-SREC 6-18
- l l ADMINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued) summarized on a quarterly basis fo'1owing the format of Appendix B thereof.
For solid wastes, the format for Table 3 in Appendix B shall be supplemented with three additional categories: class of solid wastes (as defined by 10 CFR Part 61), type of container (e.g., LSA, Type A, Type B, Large Quantity) and SOLIDIFICATION agent or absorbent (e.g., cement, urea formaldehyde).
The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, l wind direction, atmospheric stability, and precipitation (if measured), or in L the form of joint frequency distributions of wind speed, wind direction, and l atmospheric stability.* This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall.also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside l the SITE BOUNDARY (Figure [5.1-3]) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time, and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous efflu-ents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall i
be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).
The Semiannual Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall also include an assessment of radia-tion doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary efflu-ent pathways and direct radiation, for the previous calendar year to show con-formance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribu-tion from liquid and gaseous effluents are given in Regulatory Guide 1.109, i Rev. 2, October 1977.
. The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.
The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP)
- In lieu of submission with the Semiannual Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.
GE-SREC 6-19 L-___-_---_______-__________ .
DRAFT ADMINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued) and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), pursuant to Control 6.14, as well as~any major change to Liquid, Gaseous, or Solid Radwaste Treatment Systems. It shall also include a listing of new locations for dose calcu-
-lations and/or environmental monitorirg identified by the Land Use Census pursuant to Control 3.12.2.
The Semiannual Radioactive Effluent Release Reports shall also include the following: an explanation as to why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Control 3.3.3.10 or 3.3.3.11, respectively; and description of the events leading to liquid holdup tanks or gas storage tanks exceeding the control limits.
l GE-SREC 6-20
.J
' ADMINISTRATIVE CONTROLS' 6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM)
Changes to the.0DCM:
- a. Shall be documented and records of reviews performed shall be retained as required by Specification 6.10.30. This documentation shall contain:
- 1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and
- 2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
- b. Shall becore effective after review and acceptance by the [URG) and the approval of the Plent Manager,
- c. Shall be submitted to the Commission in the form of a complete.
legible copy of the entire ODCM as part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.
GE-SREC 6-24
)
i-APPENDIX A Radiological Assessinent Branch Technical Position, Revision 1, Nevernber 1979 l
l l
' ReviFion 1 November 1979 Branch Technical Position
. Background
4
- Regulatory Guide 4.8, Environmental Technical Specifications for. Nuclear Power Plants, issued for comment in December 1975, is being revised based on comments received. The Radiological Assessment Branch issued a Branch Position on the radiological portion of the environmental monitoring program in March,1978.
The posit 1on was formulaten by an NRC working group which considered comments received after the issuance of the Regulatory Guide 4.8. This is Revision 1 of that Branch Position paper. The changas are marked by a vertical line in the right margin. The most significant change is the increase in direct radiation measurement stations.
10 CFR Parts 20 and 50 require that radiological environmental monitoring i programs be established to provide data on measurable levels of radiation and radioactive materials in the site environs. In addition, Appendix I to 10 CFR Part 50 requires that the relationship between quantities of radioactive material released in effluents during normal operation, including anticipated operational occurrences, and resultant radiation doses to individuals from principals pathways of exposure be evaluated. These programs should be con-ducted to verify the effectiveness of in-plant measures used for controlling the release of radioactive materials. Surveillance should be established to identify changes in the use of unrestricted areas (e.g., for agricultrual purposes) to provide a basis for modifications in the monitoring programs for evaluating doses to individuals from principal pathways of exposure. NRC Regulatory Guide 4.1, Rev.1, " Programs for Monitoring Radioactivity in the Environs of Nuclear Power Plants," provides an acceptable basis for the design of programs to monitor levels of radiation and radioactivity in the station environs.
This position sets forth an example of an acceptable minimum radiological monitoring program. Local site characteristics must be examined to determine if pathways not covered by this guide may significantly contribute to an individual's dose and should be included in the sampling program.
l
2 AN ACCEPTABLE RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM Prooram Requirements Environment.a1 samples shall be collected and analyzed according to Table 1 at I locations shown in Figure 1.1 Analytical techniques used shall be such that
.the detection capabilities in Table 2 are achieved.
The results of the radiological environmental monitoring'are intended to supf ement the results of the radiological effluent monitoring by verifying that The measurable concentrations of radioactive materials and levels of-radiation are not higher than expected on the basis of the effluent measure-ments and modeling of the environmental exposure pathways. Thus, the specified environmental monitoring program provides measurements of radiation and of radio-active materials in those exposure pathways and for tt.ose radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation. The initial radiological environmental monitoring program should be conducted for the first three years of commercial operation (or other period corresponding to a maximum burnup in the initial core cycle). Following this period, program changes may be proposed based on operational experience.
The specified detection' capabilities are state-of-the-art for routine environ-mental measurements in industrial laboratories.
Deviations are permitted from the required sampling schedule if specimens are unobtainable due to hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sempling period. All deviations from the sampling schedule shall be documented in the annual report. _
The laboratories of the licensee and licensee's contractors which perfom analyses shall participate in the Environmental Protection Agency's (EPA's)
Environmental Radioactivity Laboratory Intercomparisons Studies (Crosscheck)
Program or equivalent program. This participation shall include all of the determinations (sample medium-radionuclides combination) that are offered by EPA and that also are included in the monitoring program. The results of analysis of these crosscheck samples shall be included in the annual report.
The participants in the EPA crosscheck program may provide their EPA program code so that the NP.C can review the EPA's participant data directly in lieu of submission in the annual report.
'It may be necessary to require special studies on a case-by-case and site specific basis to establish the~ relationship between quantities of radioactive material released in effluents, the concentrations in environmental media, and the resultant doses for important pathways.
l l 3
If the results of a determination in the EPA crosscheck program (or equivalent program) are outside the speci'ied control limits, the laboratory shall inves-tigate the cause of the problem and take steps to correct it. The results cf this investigation and corrective action shall be included in the annual report.
The requirement for the participation in the EPA crosscheck program, or similar program, is based on the need for independent checks on the precision and accuracy of the measurements of radioactive material in environnertal sample matrices as part of the quality assurance program for environment? monitoring in order to demonstrate that the results are reasonably valid.
A census shall he conducted annually during the growing season to determine the location of the nearest milk animal and nearest garden greater than 50 square meters (500 sq. ft.) producing broad lear vegetation in each of the 16 meteorological sectors within a distance of 8 km (5 miles).8 For elevated releases as defined in Regulatory Guide 1.111, Rev.1. , the census shall also
, identify the locations of all milk animals, and gardens greater than 50 square meters producing broad leaT" vegetation out to a distance of 5 km. (3 miles) for each radial sector.
If it is learned from this census that the milk animals or gardens are present at a location which yields a calculated thyroid dose greater than those previously sampled, or if the census results in changes in the location used in the radioactive effluent technical specifications for dose calculations, a written report shall be submitted to the Director of Operating Reactors, NRR (with a copy to the Director of the NRC Regional Office) within 30 days identifying the new location (distance and direction). Milk animal or garden locations resulting in higher calculated doses shall be added to the surveillance program as soon as practicable.
The sampling location (excluding the control semple location) having the lowest calculated dose may then be dropped from the surveillance program at tne end of the grazing or growing season during which the census was con-ducted. Any location from whien milk can no longer be obtained may be dropped from the surveillance program after notifying the NRC in writing that they are no longer obtainable at that location. The results of the land-use census shall be reported in the annual report.
The census of milk animals and gardens producing broad leaf vegetation is based on the requirement in Appendix ! of 10 CFR Part 50 to " Identify changes in the use of unrestricted areas (e.g., for agricultural purposes) to permit modifications in monitoring programs for evaluating doses to individuals from principal pathways of exposure '- The consumption of milk from animals grazing on contaminated pasture and of la.fy vegetation contaminated by airborne a
Broad leaf vegetation sampling r.sy be performed at the site boundary in a sector with the highest D/Q in i .eu of the garden census.
- _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ ._ o
4 radiciodine is a major potential source of exposure. Samples from atik animals are considered a better indicator of radiciodine in the environment than vegetation. If the census reveals milk animals are not present or are unavailable for sampling, then vegetation must be sampled.
The 50 square meter garden, considering 20% used for growing broad leaf vegetation 2
(i.e., similar to lettuce and cabbage), and a vegetation yield of 2 kg/m ,
will produce the 26 kg/yr assumed in Regulatory Guide 1.109. Rev 1., for child consumption of leafy vegetation. The option to consider the garden to be broad leaf vegetation at the site boundary in a sector with the highest D/Q should be conservative and that location may be used to calculate doses due to radioactive effluent releases in place of the actual locations which would be determined by the census. This option does not apply to plants with elevated releases as cefined in Regulatory Guide 1.111, Rev.1.
The increase in the number of direct radiation stations is to better characterize the individual exposure (mrem) and population exposurv (man-rem) in accordance
~
with Criterion 64 - Monitoring radioactivity re1 eases, of 10 CFR Part 50, Appendix A. The NRC will place a similar amount of stations N the area between the two rings designated in Table 1.
NOTE Guidance on the subjects contained on Assessment Branch Technical Position RAB-BTP) (pages has been 4 through modified 16 of the Radiologic and upgraded based on operating cxperience since Revision 1 was published in 19'/9. The current staff guidance for the following items has been incorporated in the Section 3/4-12 and Section 6 Controls of NUREG-1301 and 1302.
Reporting Requirement
- Table 1: Operational Radiological Environmental Monitoring Report
- Table 2: Detection capabilities for Environmental Sample Analysis
- Table 4: Reporting Levels for Radioactivity Concentrations in Environmen-tal Samples The following items remain unchanged:
i
- Footnote to Table 1 on page 10 !
- Table 3 of page 14 l
- Figure 1 of page 16 l
1
l l l Pages 5, 6, 7, 8, 9, 11, 12, 13, 15 f
)
The above pages have been superceded by text and tables in fiUREG-1301 and 1302.
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I APPENDIX B
" Appendix B - General Contents of the Offsite Dose Calculation Manual (00CM) (Revision 1, February 1979)" to the paper authored by C. A. Willis and F. J. Congel, " Status of NRC Radiological Effluent Technical Specification Activities" presented at the Atomic Industrial Forum Conference I on HEPA-and Nuclear Regulation, October 4-7, 1981, Washington, D.C.
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,,. .c APPENDIX B GENERAL CONTENTS OF THE OFFSITE DOSE CALCULATION MANUAL (ODCM*)
(Rev.1. February 1979)
Section 1 - Set Points Prwide the equations and methodology to be used at the station or unit for each alarm and trip set point on each effluent release point according to the Specifications 3.3.3.8 and 3.3.3.9. The instrumentation for. each alarm and trip set point, including radiation monitoring and sampling systems and
. effluent control features, should be identified by reference to t ehFSAR (or Final Hazard Summary). This infomation should be consistent with the recommendations of Section 1 of Standard Review Plan 11.5, NUREG-75/087, (Revision 1). . If the alarm and/or trip set point value is variable, prwide
.the equation to determine the set point value to be used, based on actual.
release conditions,.that will assure that the Specification is met at each release point; and prwide the value to.be used when releases are not in progress. ' If dilution or dispersion is used, state the onsite equipment and measurement method used during release, the site related parameters and the set points used to. assure that the Specification is met at each release point. The fixed and variable set points should consider the radioactive effluent to have a radionuclides distribution represented by nomal and anticipated operational occurrences.
Section 2 - Liquid Effluent Concentration Prwide the. equations and methodology to be used at the station or unit for each liquid release point according to the Specification 3.11.1.1. For systems with continuous or batch releases, and for systems designed to monitor and control both continuous and batch releases, prwide the assump-
' tions and parameters to be used to compare the output of the monitor with the liquid concentration specified. State the limitations for combined discharges to the same release point. In addition, describe the method and assumptions for obtaining representative samples from each batch and use of previous post-release analyses or composite sample analyses to meet the Specification.
S9ction 3 - Gaseous Effluent Dose Rate Prwide the equations and methodology to be used at the station or unit for each gaseous release point according to Specification 3.11.2.1. Consider the various pathways, release point elevations, site related parameters and radionuclides contribution to the dose impact limitation. Prwide the
- The femat for the ODCM is lef t up to the licensee and may be simplified by tables and grid printout. Each page should be numbered and indicate the facility approval and effective date. .
dose factors to be used for the identified radionuclides released. Provide the annual average dispersion values (X/Q and D/Q), the site specific para-meters and release point elevations.
Section 4 - Liquid Effluent Dose Provide the equations and methodology to be used at the station or unit for each liquid release point according to the dose objectives given in Spect-fication 3.11.1.2. The section should describe how the dose contributions are to be calculated for the various pathways and release points, the equa-tions and assumptions to be used, the site specific parameters to be measured and used, the receptor location by direction and distance, and the method of estimating and updating cumulative doses due to liquid releases. The dose factors, pathway transfer factors, pathway usage factors, and dilution fac-tors for the points of pathway origin, etc., should be given, as well as receptor age group, water and food consumption rate and other factors assumed or measured. Provide the method of detemining the dilution factor
. at the discharge during any liquid effluent release and any site specific parameters used in these determinations.
Section 5 - Gaseous Effluent Dose Provide the equations and methodology to be used at the station or unit for each gaseous release. point according to the dose objectives given in Specifications 3.11.2.2 and 3.11.2.3. The section should describe how the dose contributions are to be calculated for the various pathways and release points, the equations and assumptions to be used, the site specific parameters to be measured and used, the receptor location by direction and distance, and the method to be used for estimating and updating cumulative doses due to gaseous releases. The location, direction and distance to the nearest resi-dence, cow, goat, meat animal, garden, etc., should be given, as well as receptor age group, crop yield, grazing time and other factors. assumed or measured. Provide the method of detemining dispersion values (X/Q and D/Q) for releases and any site specific parameters and release point elevations used in these determinations.
Section 6 - Projected Doses For liquid and gaseous radwaste treatment systems, provide the method of projecting doses due to effluent releases for the normal and alternate pathways of treatment according to the specifications, describing the com-ponents and subsystems to be used, i
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Section 7 - Operability of Eouipment Provide a flow diagram (s) defining the treatment paths and the components of the radioactive liquid, gaseous and solid waste management systems that are to be maintained and used, pursuant to 10 CFR 50.36a, to meet Technical Specifications 3.11.1.3, 3.11.2.4 and 3.11.3.1. Subcomponents of packaged equipment can be identified by a list. For operating reactors whose con-struction permit applications were filed prior to January 2,1971, the flow diagram (s) shall be consistent with the information prwided in conformance with Section V.B.1 of Appendix I to 10 CFR Part 50. For OL applications whose construction permits were filed efter January 2,1971, the flow diagram (s) shall be consistent with the information provided in Chapter 11 .
of the Final Safety Analysis Report (FSAR) or amendments thereto.
Section 8 - Sample Locations Provide a map of the Radiological Environmental Monitoring Sample Locations indicating the numbered sampling locations given in Table 3.12-1. Further clarification on these numbered sampling locations can be prwided by a list, indicating the direction and distance from the center of the buil. ding com-plex of the unit or station, and may include a descriptive name for identi-fication purposes.
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1 APPENDIX C GENERIC LETTER 89-01 IMPLEMENTATION OF PROGRAMMATIC CONTROLS FOR RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS 'IN THE ADMINISTRATIVE. CONTROLS SECTION OF .
THE TECHNICAL SPECIFICATIONS AND THE RELOCATION OF PROCEDURAL DETAILS OF RETS TO THE OFFSITE DOSE CALCULATION MANUAL OR TO THE PROCESS CONTROL PROGRAM-
/ %, UNITED STATES
!" n NUCLEAR REGULATORY COMMISSION '
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, wAssiwovow, o. c.20sss o.. o January 31, 1989 i
TO ALL POWER REACTOR LICENSEES AND APPLICANTS
SUBJECT:
IMPLEME!JTATION OF PROGRAMMATIC CONTROLS FOR RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS IN THE ADMINISTRATIVE CONTROLS SECTION OF THE TECHNICAL SPECIFICATIONS AND THE RELOCATION OF PROCEDURAL DETAILS OF RETS TO THi 0FFSITE DOSE CALCULATION XANUAL OR TO THE PROCESS CONTROL PROGRAM (GENERIC LETTER 89-01)
The NRC staff has examined the contents of the Radiological Effluent Technical Specifications (RETS) in relation to the Comission's Interim Policy Statement on Technical Specification Improvements. The staff has determined that pro-gramatic controls the Technical can be innp(lemented Specifications TS) to satisfyinexisting the Administrative Controls section of regulatory requirements for RETS. At the same time, the procedural details of the current TS on radio-active effluents and radiological environmental monitoring can be relocated to theOffsiteDoseCalculationManual(0DCM). Likewise, the procedural details of the current TS on solid radioactive wastes can be relocated to the Process Control Program (PCP). These actions simplify the RETS, meet the regulatory reovirements for radioactive effluents and radiological environmental monitor-ing, and are provided as a line-item improvement of the TS, consistent with the goals of the Policy Statement.
New programmatic controls for radioactive effluents and radiological environ-mental monitoring are incorporated in the TS to conform to the regulatory requirements of 10 CFR 20.106, 40 CFR Part 190,10 CFR 50.36a, and Appendix 1 to 10 CFR Per0 50. Existing programmatic reoutrements for the PCP are being retained in the TS. The procedural details included in licensees' present TS cn radioactive effluents, solid radioactive wastes, environmental monitoring, and associated reporting reovirements will be relocated to the ODCM or PCP as appropriate. Licensees will handle future changes to these procedural details in the ODCM and the PCP under the administrative controls for changes to the ODCM or PCP. Finally, the definitions of the ODCM and PCP are updated to reflect these changes.
Enclosure 1 provides guidance for the preparation of a license amendment re-ouest to implement these alternatives for RETS. Enclosure 2 provides a list-irg of existing RETS and a description of how each is addressed. Enclosure 3 prevides model TS for programmatic controls for RE15 and its associated report-ing recu1rements. Finally, Enclosure 4 provides model specifications for retaining existing requirements for explosive gas monitoring instrumentation requirements that apply on a plant-specific basis. Licensees are encouraged to propose changes to TS that are consistent with the guidance provided in the enclosures. Conforming ac.endment reouests will be expeditiously reviewed by
1
- ." Ce'neric tr,tter 89-01' 2- January 31,1989 j
the NRC Project Manager for the facility. Proposed amendments that deviate:
from this guidance will require a longer, more detailed review. Please contact the appropriate Project Manager if you have questions on this matter.
Sincerely, Ya Acting Associate D ctor for Projects Office of Nuclear Reactor Regulation "
Enclosures:
I through 4 as stated l
l I
i l Generic Lett;r 89 01 ENCLOSURE 1 )
l GUIDANCE FOR THE IMPLEMENTATION OF PROGRAMMATIC CONTROLS FOR RETS IN THE ADMINISTRATIVE CONTROLS SECTION OF TECHNICAL SPECIFICATIONS j AND THE RELOCATION OF PROCEDURAL DETAILS OF CURRENT RETS TO THE OFFSITE DOSE CALCULATION MANUAL OR PROCESS CONTROL PROGRAM INTRODUCTION This enclosure provides guidance for the preparation of a license amendment request to implement programmatic controls in Technical Specifications (TS) i for radioactive effluents and for radiological environmental monitoring con-forming to tha applicable regulatory requirements. This will allow the reloca-tion of existing procedural details of the current Radiological Effluent Technical Specifications (RETS) to the Offsite Dose Calculation Manual (ODCM).
Procedural details for solid radioactive wastes will be relocated to the ;
Process Control Program (PCP). A proposed amendment will (1) incorporate pro- '
grammatic controls in the Administrative Controls section of the TS that sat- j isfy the requirements of 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a. and '
Appendix I to 10 CFR Part 50, (2) relocate the existing procedural details in current specifications involving radioactive effluent monitoring instruments-tion, the control of liquid and gaseous effluents, equipment requirements for liquio and gaseous effluents, radiological environmental monitoring, and radio-logical reporting details from the TS to the ODCM, (3) relocate the definition of solidification and existing procedural details in the current specification on solid radioactive wastes to the PCP, (4) simplify the associated reporting requirements, (5) simplify the administrative controls for changes to the ODCM and PCP, (6) add record retention requirements for changes to the ODCM and PCP, and (7) update the definitions of the ODCM and PCP consistent with these -
changes. 1 The NRC staff's intent in recommending these changes to the TS and the reloca-tion of procedural details of the current RETS to the ODCM and PCP is to ful-fill the goal of the Commission Policy Statement for Technical Specification Improvements. It is not the staff's intent to reduce the level of radiological effluent control. Rather, this amendment will provide programmatic controls for RETS consistent with regulatory requirements and allow relocation of the l procedural details of current RETS to the ODCM or PCP. Therefore, future I changes to these procedural details will be controlled by the controls for changes to the ODCM or PCP included in the Administrative Controls section of the TS. These procedural details are not required to be included in TS by 10 CFR 50.36a.
]
DISCUSSION Enclosure 2 to Generic Letter 89- provides a summary listing of specifica-tions that are included under the heading of RETS in the Standard Technical l Specifications (STS) and their disposition. Most of these specifications will l be addressed by programmatic controls in the Administrative Controls section of )'
' the TS. Some specifications under the heading of RETS are not covered by the new programmatic controls and will be retained as requirements in the existing plant TS. Examples include requirements for explosive gas monitoring instru- )
mentation, limitations on the quantity of radioactivity in liquid or gaseous 1 holdup or storage tanks or in the condenser exhaust for BWRs, or limitations on '
)
explosive gas mixtures in offgas treatment systems and storage tanks.
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g G.eneric Lstter 89- 01 Enclosure l' Licensees with nonstandard TS should follow the guidance provided in Enclo-sure 2 for the disposition of similar requirements in the format of their TS.
Because solid radioactive wastes = are addressed under existing programmatic controls for the Precess Control Program, which is a separate program from the new programmatic controls for' liquid and gaseous- radioactive affluents the requirements for solid radioactive wastes and associated solid waste reporting requirements in current TS are' included as procedural details that will be .
relocated to the PCP as part of this line-item improvement of TS, Also, the
. staff has concluded that records of licensee reviews performed for changes made to the ODCM and PCP should be documented and retained for the duration of the unit operating license. This approach is in lieu of the current requirements that the reasons' for changes to the ODCM and PCP be addressed in the Semiannual Effluent Release Report.
The following items are to be included in a license amendment request to imple-ment these changes. - First, the model specifications in Enclosure 3 to Generic Letter 89- should be incorporated into the TS to satisfy the requirements of 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50.
.The definitions of.the ODCM and PCP should be updated to reflect these changes.
The programmatic and reporting requirements are general in nature 'and do.not contain plant-specific details. Therefore, these changes to the Administrative Controls section of the TS are to replace corresponding requirements in plant T5'that address these items. They should be proposed for incorporation into
! the plant's TS without change in substance to replace existing requirements.
If necessary, only changes in format should be proposed. If the current TS include requirements for explosive gas monitoring instrumentation as part of the gaseous effluent monitoring instrumentation requirements, these require-ments should be retained. Enclosure 4 to Generic Letter 89- provides model specifications for retaining such requirements. i Second, the procedural details covered in the licensee's current RETS, consist-ing of the limit , conditions for operation, their applicability, remedial actions, surveill...ce requirements, and the Bases section of the TS for these requirements, are to be relocated to the ODCM or PCP as appropriate and in a manner that ensures that these details are incorporated in plant operating pro-cedures. The NRC staff does not intend to repeat technical reviews of the re-located procedural details because their consistency with the applicable regula-tory requirements is a matter of record from past NRC reviews of RETS. If licensees make other than editorial changes in the procedural details being transferred to the ODCM, each change should be identified by markings in the margin and the requirements of new Specification 6.14a.(1) and (2) followed.
Finally, licensees should confirm in the amendment request that changes for relocating the procedural details of current RETS to either the ODCM or PCP have been prepared in accordance with the proposed changes to the Administra-tive Controls section of the TS so that they may be implemented immediately upon issuance of the proposed amendment. A complete and legible copy of the revised ODCM should be forwarded with the amendment request for NRC use as a reference. The NRC staf f will not concur in or approve the revised ODCM.
Ceneric Letter'89-01 Enclosure 1 Licensees should refer to "Generi- Letter 89- " in the Sub.iect line of license amendment reauests implementing tee guidance of this Generic Letter. This will facilitate the staff's tracking of licensees' responses to this Generic Letter.
SUMMARY
The license amendment recuest for the line-item improvements of the TS relative to the RETS will entail (1) the incorporation of programmatic controls for-radioactive ~ effluents and radiological environmental monitoring in the Admin-istrative Controls section of the TS, (2) incorporatation of the procedural details of the current RETS in the ODCM or PCP as appropriate, and (3) confirm-ation that the guidance of this Generic Letter has been followed. .
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TECHNICAL SPECIFICATIONS TO BE REVISED l 1.17 DEFINITIONS: OFFSITE DOSE CALCULATION MANUAL.
i 1.22 DEFINITIONS: PROCESS CONTROL PROGRAM-6.8.4 g. PROCEDURES AND PROGRAMS: RADIOACTIVE EFFLUENT CONTROLS 6.8.4 h. PROCEDURES AND PROGRAMS: RADIOLOGICAL ENVIRONMENTAL MONITORING 6.9.1.3 REPORTING REQUIREMENTS: ANNUAL RADIOLOGICAL ENVIRONMENTAL DPERATING REPORT 6.9.1.4 REPORTING REQUIREMENTS: SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 6.10 RECORD RETENTION 6.13 PROCESS CONTROL PROGRAM (PCP) 6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM)
MODEL TECHNICAL SPECIFICATION REVISIONS (To supplement or replace existing specifications) 1.0 DEFINITIONS .
OFFSITE DOSE CALCULATION MANUAL 1.17- The OFFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radio-active gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alam/ Trip Setpoints, and in the conduct of the Environ-cental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Pro-grams required by Section 6.8.4 and (2) descriptions of the infomation that should be included in the Annual Radiological Environmental Operating and Semi-annual Radioactive Effluent Release Reports required by Specifications 6.9.1.3 and 6.9.1.4.
1.22 :The PROCESS CONTROL PROGRAM (PCP) shall con +.ain the current fomulas, sampling, analyses, test, and determinations to be made to ensure that process-ing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure complir.nce with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste. -
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i i .'- Gbneric L;tter 89- 01 Enclesure 3 l
6.0 ADMINISTRATIVE' CONTROLS 6.8 PROCEDURES AND PROGRAMS 6.8.4 The following programs shall be established, implemented, and maintained:
- g. Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to MEMBERS OF THE PUBLIC from ~ radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shell in-clude remedial actions to be taken whenever the program limits are q exceeded. The program shall include the following elements: l
- 1) Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and set-point determination in accordance with the methodology in the 1 ODCM,
- 2) Limitations on the concentrations of radioactive material released in liquid effluents to UNRESTRICTED AREAS conforming to 10 CFR Part 20, Appendix B. Table II, Column 2,
- 3) Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.106 and with the methodology and parameters in the ODCM,
- 4) Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conforn.-
ing to Appendix I to 10 CFR Part 50,
- 5) Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days,
- 6) Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radio-activity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50,
- 7) Limitations on the dose rate resulting from radioactive material released in gaseous effluents to area's beyond the SITE BOUNDARY conforming to the doses associated with 10 CFR Part 20, Appendix B, Table II, Column 1,
i Z C4neric Letter 8941 Enclosure 3' l 3
ADh '51RATIVE CONTROLS
- 6.8.4 g. Radioactive Effluent Controls Program (Cont.) j
- 8) Limitations on the annual and quarterly air doses resulting from noble gases. released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,
{
- 9) Limitations on the annual and quarterly doses to a MEMBER OF .
THE PUBLIC from Iodine-131,-Iodine-133, tritium, and all radio-nuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the -
SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50, i
, 10) Limitations on venting and purging of the Mark II containment through the Standby Gas Treatment System to maintain releases 4 as low as reasonably achievable (BWRs w/ Mark II containments), '
and
- 11) Limitations on the annual dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 190.
- h. Radiological Environmental Monitorino Program A program shall be provided to monitor the radiation and radio-nuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental expo-sure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:
- 1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the method-
. ology and parameters in the ODCM,
- 2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifica- i tions to the monitoring program are made if required by the results of this census, and
- 3) Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance pro-gram for environmental monitoring.
. _ - - - . _ - - - _ _ _ - _ . . - - _ . - - _ _ . - _ - _ - - _ _ - - _ - _ _ . - - _ . _ _ _ - - - - - - - . _ . - - _ _ - - _ _ . - - . _ - -_. - - - - _ _ . . _ _ . . - _ _ _ . - - - __.-)
.- G;neric LGtter 89-01 Enc 1csure 3'
-ADMINISTRATIVE CONTROLS E.9 REPORTING REQUIREMENTS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT" l
6.9.1.3 The Annual Radiological Environmental Operating Report covering the
. operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpreta-
. tions, and analysis of t: ends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be -
consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.
SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT **
6.9.1.4 The Semiannual Radioactive Effluent Release Report cover'ing the oper-ation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The report shall in-clude a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be
. (1) consistent with the objectives outlined in the ODCM and PCP and (2) in con-formance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.
6.10 RECORD RETENTION 6.10.3 The following records shall be retained for the duration of the unit Operating License:
- o. Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.
6.13 PROCESS CONTROL PROGRAM (PCP)
Changes to the PCP:
- a. Shall be documented and records of reviews performed shall be retain-ed as required by Specification 6.10.30. This documentation shall contain:
- 1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and
- A single submittal may be made for a multi-unit station.
- A single submittal may be made for a multi-unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
- Generic Lctter 89-01 Enc 1ssure 3 l ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM (PCP) (Cont.)
- 2) A determination that the change will maintain the overall con-formance of the solidified waste product to existing require-ments of Federal State, or other applicable regulations.
- b. Shall become effective after review and acceptance by the [URG) and i the approval of the Plant Manager.
l 6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM)
Changes to the ODCM:
- a. Shall be documented and records of reviews performed shall be retain-ed as required by Specification 6.10.30. This documentation shall contain:
- 1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and
- 2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.
- b. Shall become effective after review and acceptance by the [URG) and the approval of the Plant Manager.
. c. Shall be submitted to the Commission in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Semiannual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., sonth/ year) the change was implemented. -
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. Generic Lett:r B9- 01 Enc 1:sure'4 MODIFICATION OF THE SPECIFICATION FOR RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION TO RETAIN REQUIREMENTS FOR EXPLOSIVE GAS MONITORING INSTRUMENTATION ,
INSTRUMENTATION EXPLOSIVE RABIBAEliVE GASEBBS-EFFtWENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION explosive 3.3.3.11 The radienetive gaseess-efficent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their Alarm / Trip Setpoints set to 01sure that the limits of Specifications-3 ll:E:1-and 3.11.2.5 are not Cxceeded. The-Aia*m/ Trip-Setpeints-ef-these-channeis meeting-Specification S:11-E-1-shali-be-determined-and-edjested-in-accordance-with-the-methodelegy cnd paramtters-in-the-8BEM:
APPLICABILITY: As shown in Table 3.3-13 ACTION: ,
explosive
- a. With an radiesetive gaseens-efficent monitoring instrumentation channeT Alarm / Trip Setpoint less conservative than required by the above specification--immediately-suspend-the-release-ef-radienetive gaseens-efficents-menitored-by-the-affected-channei--or declare the channel inoperable and take the ACTION shown in Table 3.3-13.
explosive
- b. With less than the minimum number of radienetive gaseens-efficent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-33. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful explain-in-the-next-Semi-annesi-Radienetive-Efficent-Reiesse-Repert prepare and submit a Special Report to the Commission pursuant to Specification 6:9:i:4 6.9.2 to explain why this inoperability was not corrected in a timely manner.
- c. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS i
explosive l 4.3.3.11 Each radienetive gaseens-efficent monitoring instrumentation channel I shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOUREE EHEEK- CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-9.
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ACTION STATEMENTS ACTION 45 - (Not used)
- i. ACTION 46 - (Not used)
ACTION 47 - (Not used)
ACTION 48 . (Not used)'
ACTION 49 - . With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of this WASTE GAS HOLDUP SYSTEM may continue provided grab samples are collected at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and analyzed within the following 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
ACTION 50 - With the' number cf r:hannels OPERABLE one less than required by.
the Minimum Channels OPERABLE requirement, operation of this system may continue provided grab samples are taken and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With both channels inoperable, operation may continue provided grab samples are taken and analyzed at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during degassing operations and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> daring other operations.
ACTION 51 - (Not used)
ACTION 52 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend oxygen supply to the recombiner.
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- . , .Gqneric L;tter _89- 01 p
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TABLE 4.3-9 (Continued)
TABLE NOTATIONS
'O (Not used).
C* During~ WASTE. GAS HOLDUP SYSTEM operation.
(1) '(Not used)
(2) (Not used)
D - (3) (Not used)
(4) . The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
- a. One volume percent hydogen, balance nitrogen, and
(5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
2 Sample STS 3/4 3-(n+4)
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- - E LIST OF RECENTLY ISSUED GENERIC LETTERS 1
Generic Date of -
Letter No. Sub.iect Issuance Issued To 88-20 INDIVIDUAL PLANT - 11/23/88 ALL LICENSEES HOLDING EXAMINATION FOR SEVERE OPERATING LICENSES ACCIDENT VULNERABILITIES - AND CONSTRUCTION 10 CFR 50.54(f) PERMITS FOR NUCLEAR POWER REACTOR FACILITIES 68-19 USE OF DEADLY FORCE BY 10/28/88 ALL FUEL CYCLE FACILITY LICENSEE GUARDS TO PREVENT LICENSEES WHO POSSESS.
THEFT OF SPECIAL NUCLEAR USE. IMPORT EXPORT, MATERIAL . OR TRANSPORT FORMULA QUANTITIES OF STRATEGIC SPECIAL NUCLEAR MATERIAL 88-18 PLANT RCCORD FJORAGE ON 10/20/88 ALL LICENSEES OF OPTICAL DISKS -OPERATING REACTORS AND HOLDERS OF CONSTRUCTION PERMITS 88-17 LOSS OF DECAY HEAT REMOVAL 10/17/88 ALL HOLDERS OF 10 CFR 50.54(f) OPERATING LICENSES OR CONSTRUCTION PERMITS FOR PRES $URIZED WATER REACTORS 88-16 REMOVAL OF CYCLE-SPECIFIC 10/04/88 ALL POWER REACTOR PARAMETER LIMITS FROM LICENSEES AND TECHNICAL SPECIFICATIONS APPLICANis 88-15 ELECTRIC POWER SYSTEMS - 09/12/88 ALL POWER REACTOR INADEQUATE CONTROL OVER LICENSEES AND DESIGN PROCESSES APPLICANTS 88-14 INSTRUKENT AIR SUPPLY 08/08/88 ALL HOLDERS OF __
SYSTEM PROBLEMS AFFECTING OPERATING LICENSES SAFETY-RELATED EQUIPMENT OR CONSTRUCTION PERMITS FOR NUCLEAR POWER REACTORS 88-13 OPERATOR LICENSING 08/08/88 ALL POWER REACTOR EXAMINATIONS LICENSEES AND APPLICANTS FOR AN OPERATING LICENSE.
88-12 REMOVAL OF FIRE PROTECTION 08/02/88 ALL POWER REACTOR REQUIREMENTS FROM TECHNICAL LICENSEES AND SPECIFICATIONS APPLICANTS
- _ - - - - _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ , _