ML20092D830
| ML20092D830 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 06/19/1984 |
| From: | MISSISSIPPI POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20092D825 | List: |
| References | |
| NUDOCS 8406220164 | |
| Download: ML20092D830 (100) | |
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{{#Wiki_filter:, 22: 3
- ATTACHMENT TO AECM-84/0338 (6/19/84)
PROPOSED CHANGES TO THE GRAND GULF NUCLEAR STATION TECHNICAL SPECIFICATIONS NRC TECHNICAL REVIEW BRANCH: RADIOLOGICAL ASSESSMENT O B406220164 840619 PDR ADOCK 05000416 P PDR
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Atttchesnt / Radiological AsssssAsnt AECM-84/0338 (6/19/84) Page 2 Listing of Item Numbers by Technical Specification Problem Sheet (TSPS) Namber TSPS No. Item Nos.* , 036 C.01 085 D.01 e 086 D.02 088 D.03 089 D.04 090 D.05 091 A.01 092 D.06 105 A.04 - 190 D.07 191 D.08 192 D.09 ~- 194 D.10 225 D.11 248 A.03 249 D.12, A.02 .i w-
*1 tem number format: A.02 Item number within category-Category designator 268mm2
~ ' Attcchssnt / Rrdiological Asssesesnt AECM-84/0338 (6/19/84)
Psge 3 l A. TYPOGRAPHICAL ERRORS, EDITORIAL CHANGES, AND CLARIFICATIONS ) These[roposedchangescorrectobvioustypographicalerrors, implement ; editorial changes such as correction of spelling errors, punctuation
'rrors, and grammatical errors or provide clarification of the basic meaning or intent of the subject technical specifications.
HP&L has determined that the proposed changes do not: o Involve a significant increase in the probability or consequences of an accident previously evaluated; or o Create the possibility of a new or different kind of accident from any accident previously evaluated; or . o Involve a significant reduction in a margin of safety. Therefore, the proposed changes do not involve a significant hazards consideration. Several of the technical specification changes associated with this i submittal are being proposed in order to provide consistency with NUREG-0473 Rev. 3, Draf t 7 . In addition, these changes have been evaluated to assure that the changes are consistent with the present Grand Gulf programs and policies concerning radiological control and releases. These changes have been determined to be consistent with the NRC guidance and the standards for protection against radiation as specified in 10 CFR Paet 20. A description of these changes including necessary justification for the changes is provided below: TYPOGRAPHICAL ERRORS 4 Typographical errors are being corrected by this submittal as listed below. Correction of these typographical errors is purely an administra-tive change. (See attached revised technical specification pages for exact changes proposed.) TSPS No. TS Page No.
- 1. 91 1-3 3/4 12-11
'2 . 249 B 3/4 11-2 3/4 12-1 3/4 12-3 3/4 12-5 l
6-25 l l 268mm3 . 2 - - - . _ . _ - ._ m __ _ . , _ - ._
Attechnent / Radiological Assescarnt AECM-84/0338 (6/19/84) Page 4 EDITORIAL CHANGES _ l A prop sed editorial changes to the technical specifications is discussed } below: - .
- 3. (TSPS 248) Renumbering of Bases, Technical Specification Bases
- 3/4.11.2.4 and 3/4.11.2.5 This proposed change renumbers Bases 3/4.11.2.4 and 3/4.11.2.5 to reflect the technical specification sections described in the text of the bases. This is an editorial change only and is therefore purely administrative in nature. (Page B 3/4 11-4)
CLARIFICATIONS
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Clarifications to the technical specifications to improve understanding and readability are discussed below:
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- 4. (TSPS 105), SITE BOUNDARY Terminology, Technical Specifications 1.40, 3/4.11.2.1, 3/4.11.2.3, 3/4.11.2.5, 5.1.3, 6.9.1.9, Table 3.12.1-1 and Bases 3/4.11.2.1, 3/4.11.2.2, 3/4.11.2.3, 3/4.11.2.7, 3/4.12.2 A revision is proposed to add a definition for SITE BOUNDARY to the
' technical specifications to indicate that this is the line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. This definition is being added to clarify the boundary used for offsite dose calculations. This change also adds this defined term where applicable in the technical specifications. This change does not adversely impact plant safety because it serves only to clarify the subject specifications. A revision to Technical Specification 5.1.3 is requested to change the " UNRESTRICTED AREA BOUNDARY" terminology in this design feature summary to " UNRESTRICTED AREA and SITE BOUNDARY." This proposed change will achieve consistency throughout the technical specifica-tions with the new definition for UNRESTRICTED AREA and the new- , definition far SITE BOUNDARY. This proposed change involves no , safety significance as it represents a purely administrative change ! of an editorial nature for the purpose of providing consistency in the technical specifications. (Pages 1-8, 3/4 11-8, 3/4 11-13, 3/4 j i 11-15, 3/4 12-3, 5-1, 6-18, B 3/4 11-2, B'3/4 11-3, B 3/4 11-4, B 3/4 11-5, B_3/4 12-1) l l t l 268mm4
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AECM-84/0338 (6/19/84) Page 5 B. TECHNICAL SPECIFICATION /AS-BUILT PLANT CONSISTENCY Notec'finicalsp'ecificationchangesinthiscategoryareincludedwith this attachment. .
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l Attachmsnt / Radiological Assassaant I AECM-84/0338 (6/19/84) Page 6 C. ENHANCEMENTS THAT ARE CONSISTENT WITH THE SAFETY ANALYSES The foY10 wing' proposed change is an enhancement which is consistent with the safety analyses and the licensing basis and which provide _s clarifica-tion, renders areas consistent with the philosophy and intent of the technical specifications, or provides additional plant operational margin. i Several of the technical specification changes associated with the items being reviewed by the. Radiological Assessment Branch are being proposed in order to provide consistency with NUREG-0473. Revision 3 Draft 7, which contains the most current industry practice and regulatory guidance. These changes have been evaluated to assure consistency with present Grand Gulf programs and policies concerning radiological control and releases. Furthermore, these changes have been determined to be consistent with the NRC guidance and the standards for protectibn against radiation as specified in 10 CFR 20. Since this proposed change is included in the current licensing bases and is bounded by existing safety analyses, the proposed change does not: I o Involve a significant increase in the probability or consequences of an accident previously evaluated; or t
- 4 o Create the possibility of a new or different kind of accident from any accident previously evaluated; or o involve a significant reduction in a margin of safety.
Therefore, the proposed change does not involve a significant hazards , consideration. J l A description of this change including justification for the change is j provided below:
- 1. (TSPS 036), Operational Radiological Environmental Monitoring Program, Technical Specification Table 3.12.1-1 The proposed change adds a requirement to perform gamma isotopic and 1-131 analysis for the Food Products entry listed in the Type and Frequency of Analysis column of Table 3.12.1-1. In addition, this
- change adds the superscript "c" to the " gamma isotopic" entries in this column.- Note e defines what is meant by " gamma isotopic analysis." These changes are considered enhancements.to safety in
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that they represent an additional requirement that is not presently included in the specification, clarify analysis requirements, and achieve consistency throughout the technical specifications. (Pages 3/4 12-4 and 3/4 12-5) 1 l l 4 e 1 268mm6
Attechmsnt / Rxdiological Assessmint' AECM-84/0338 (6/19/84) Page 7 D. REGULATORY REQUIREMENTS / REQUESTS / RECOMMENDATIONS The following changes are proposed to render the technical specifications consistent with recent changes in NRC policy and the Code of, Federal Regulations, as well as to implement changes or enhancements recently i requested or recommended by NRC reviewers. l s In particular, several of the technical specification changes associated with the items being reviewed by the Radiological Assessment Branch are being proposed in order to provide consistency with NUREG-0473 Revision 3, Draft 7, which contains the most current industry practice and regulatory guidance. These changes have been evaluated to assure consistency with present Grand Gulf programs and policies concerning radiological control and releases. Furthermore, these changes have been determined to be consistent with the standards for protection a' gainst radiation as specified in 10 CFR 20. These proposed changes are required to render the technical specifications consistent with recent NRC guidance, and it has been concluded based on a review of each item that the proposed changes do not: o involve a significant increase in the probability or consequences of an accident previously evaluated; or o Create the possibility of a new or different kind of accident from any accident previously evaluated; or o involve a significant reduction in a margin of safety. Therefore, the proposed changes do not involve a significant hazards consideration. I A description of these changes including justification for the changes is provided below:
- 1. (TSPS 085), Process Control Program and Solidification, Technical Specifications 1.32, 1.41, 3/4.11.3, and Bases 3/4.11.3 The following changes to the subject technical specifications and bases are proposed:
\ The definition for PROCESS CONTROL PROGRAM (PCP) is expanded to-i a. j include reference to 10 CFR 20, 10 CFR 61, 10 CFR 71, Federal j and State regulations and burial ground regulations governing the disposal of radioactive waste. This change is an enhancement to safety in that it will help to ensure compliance with applicable requirements governing radioactive waste - disposal. (Page 1-6) 4 4 Z68mm7 - f - .
' ' Attachesnt / Radiological Assessa nt AECM-84/0338 (6/19/84)
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- b. The definition for SOLIDIFICATION is revised to simplify the
' present definition to read "the conversion of wet wastes into a
'_ form that meets shipping and burial ground requirements." The present definition implies t, hat dewatering is not an appropri-ate means of SOLIDIFICATION; however, this method 6f SOLIDIFI-CATION is generally accepted at burial grounds. This change will not adversely impact plant safety as it serves only to clarify the definition. (Page 1-8) ! c. A revision to Technical Specification 3/4.11.3 concerning solid radioactive waste is proposed to reflect that SOLIDIFICATION will be assured by the provisions of the PROCESS CONTROL PROGRAM rather than by the OPERABILITY of the Solid Radwaste System. This revision is consistent with the revised defini-tion for PROCESS CONTROL PROGRAM (Item D.1.a above) which includes provisions to ensure that waste SOLIDIFICATION is accomplished consistent with applicable regulations. The proposed change replaces Technical Specification 3/4.11.3 with a new specification which includes a Limiting Condition for Operation, ACTION statement, and surveillance requirements which rely upon the PROCESS CONTROL PROGRAM to verify SOLIDIFI-l
~ CATION by periodically testing samples of waste batches. Therefore, the OPERABILITY requirement of the solid radwaste system can be deleted from this technical specification without adversely impacting plant safety because the PROCESS CONTROL PROGRAM itself provides an equivalent level of assurance of SOLIDIFICATION. (Pages 3/4 11-18 and 3/4 11-19)
- d. A revision to Bases 3/4.11.3 for SOLID RADIOACTIVE WASTE is proposed to delete the first sentence which indicates that the Solid Radwaste System will be OPERABLE when solid radwastes are being processed and packaged for shipment to offsite burial locations. This deletion is proposed to clarify that dewater-
! ing processes and contracted solid radwaste services are appropriate alternatives to preparing solid radwaste using the Solid Radwaste System. This change will not adversely impact l plant safety because it serves only to clarify the bases and does not affect any technical specification requirements. (Page B 3/4 11-5) These proposed changes are in response to NRC requests to incorpo-rate recent regulatory activities into the GGNS technical specifica-l tions. These proposed changes are revisions that will help to ensure compliance with regulatory requirements and thus,-do not adversely impact plant safety.
- 2. (TSPS 086), Liquid Waste Treatment System and VENTILATION EXHAUST TREATMENT SYSTEM, Technical Specifications 3/4.11.1.3, 3/4.11.2.5, i
and Bases 3/4.11.1.3 The proposed changes to the subject specifications.are as follows: l a. Revise the Limiting Condition for Operation for Technical Specifications 3.11.1.3 and 3.11.2.5 to specify that the i i 268mm8
Attachaznt / Radiologicci Asssssssnt
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Page 9 subject systems sh.all be used to reduce radioactive materials
,- in waste prior to their discharge when the projected doses due to effluent releases from each reactor unit to UNRESTRICTED
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AREAS (3.11.1.3) and areas at and beyond the SITE BOUNDARY i (3.11.2.5) would exceed specified values. This change is an enhancement to plant safety because it replaces the existing OPERABILITY requirements for specific components with a broader
- requirement for system OPERABILITY.
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- b. Add the phrase "other than when the VENTILATION EXHAUST l TREATMENT SYSTEM is undergoing routine maintenance" to the applicability statement of Technical Specification 3.11.2.5.
This change is an operational enhancement to allow for routine system maintenance. It is also considered an enhancement to safety because such maintenance is expected to result'in increased system availability.
- c. Delete the 31 days allowed for the systems to be inoperable from ACTION a of Technical Specifications 3.11.1.3 and
-3.11.2.5. This is an enhancement to safety because it reduces the time period between identification of system inoperability and entry into the appropriate ACTION statement.
- d. Revise ACTION Statements a.1 of Technical Specification 3.11.1.3 and 3.11.2.5 to include additional reporting requirements. This change is a safety enhancement in that imposes additional requirements not contained in the current specifications.
- e. Delete reference to Technical Specification 6.9.1.11 in ACTION
- b of Specifications 3.11.1.3 and 3.11.2.5 because this section has been deleted by problem sheet 093 (Special Reporting Requirements). A
- f. Revise Surveillance Requirements 4.11.1.3.1 and 4.11.2.5.1 to 4
specify that doses due to releases from each reactor unit to specified areas shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM. This change does not adversely impact plant safety as it serves only to clarify the subject surveillance requirements.
- g. Revise Surveillance Requirement 4.11.1.3.2 to read "The installed liquid radwaste system shall be demonstrated OPERABLE by meeting Technical Specifications 3.11.1.1-and 3.11.1.2."
Compliance with the revision will ensure that the liquid radwaste system is fulfilling its function and that releases are below allowable levels. This ,hange c is an enhancement to safety in that'it provides for a direct, quantitative determination of system OPERABILITY.
- h. Revise Surveillance Requirement 4.11.2.5.2 to read, "The installed VENTILATION EXHAUST TREATMENT SYSTEM shall be
- demonstrated OPERABLE by_ meeting Technical Specifications 3.11.2.1 and 3.11.2.2 or 3.11.2.3." Compliance with this 168mm9
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Page 10 revision will ensure that the VENTILATION EXHAUST TREATMENT SYSTEM is capable of performing its function. This change is T. an enhancement to safety in that it provides for a direct, quantitative determination o,f system OPERABILITY. .
- i. Bases 3/4.11.1.3 for Liquid Radwaste is revised to delete the first sentence which indicates the liquid radwaste system will be OPERABLE when liquid effluents require treatment prior to release to the environment. This change is made to reflect the revision to Technical Specification 3.11.1.3 (which deletes the requirement that the Liquid Radwaste System components as specified in the ODCM be OPERABLE, and requires only that the liquid radwaste system be capable of meeting the dose and concentration requirements specified in Specifications 3.11.1.1 and 3.11.1.2). Also, a paragraph is added to clarify that for units with a shared radwaste system, the liquid effluents are proportioned among the units sharing the system. These changes do not adversely impact plant safety because they serve only to clarify the bases and make them consistent with the revised technical specifications.
These proposed changes clarify the OPERABILITY, ACTION, Applica-bility, and surveillance requirements of the subject technical specifications and are consistent with NUREG-0473, Rev. 3, Draft 7, and the requirements of Appendices A and I to 10 CFR 50. These changes promote consistency and constitute additional requirements not presently included in the technical specifications. (Pages 3/4 11-6, 3/4 11-15 and B 3/4 11-2)
- 3. (TSPS 088), Major Changes to Radioactive Waste Treatment Systems.
Technical Specification 6.15 The proposed change requires that licensee initiated major changes to the Radioactive Waste Systems (liquid, gaseous, and solid) be reported to the Commission in the Semi-annual Radioactive Effluent Release Report instead of the Monthly Operating Report. A footnote was also added to allow the licensee to submit this information as part of the annual FSAR update. The proposed change to the reporting requirements does not adversely impact plant safety because it does not affect allowable releases as specified in 10 CFR Part 20. The proposed change is consistent with NUREC-0473, Rev. 3 Draft 7 and is considered to be an enhancement consistent with NRC guidance. (Page 6-26)
'4. (TSPS 089), Radioactive Effluents - Dose, Technical Specification 3/4.11.1.2 This proposed change revises Surveillance Requirement 4.11.1.2 to read, " 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 persmeters of the ODCM at least once per 31 days."
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- Attechnent / R:diologicci Assascssnt AECM-84/0338 (6/19/84)
Page 11 1 This proposed change specifies time periods over which the doses are cumulative and is consistent with the requirements of NUREG-0473 Rev. 3, Draf t 7 , and with Surveillance Requirements 4.11.2.2 and 4.11.2.3 of the present GGNS Technical Specifications. .This change is an enhancement to safety in that it clarifies the surveillance
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requirement by specifying time periods that are in compliance with 10 CFR 50, Appendix I. (Page 3/4 11-5)
- 5. (TSPS 090), Radioactive Effluents, Technical Specification 3/4.11.4 The proposed change replaces Technical Specification 3/4.11.4 with a new specification to enhance understanding of existing dose ;
limitations. This change adds an additional surveillance requirement to provide guidance for the determination of cumulative dose contributions, in accordance with the ODCM, due to direct , 1 radiation from the reactor and radwaste storage tanks. This change is consistent with 40 CFR 190, NRC guidelines, and NUREG-0473, Rev. 3, Draft 7 and is an enhancement to safety in that it imposes an additional requirement not contained in the current specification. I (Page 3/4 11-20)
- 6. (TSPS 092), Radiological Environmental Monitoring Interlaboratory Comparison Program, Technical Specification 3/4.12.3 The surveillance requirement for the subject technical specification :
presently specifies that the participants in the Environmental Protection Agency (EPA) crosscheck program may provide the EPA program code for NRC review in lieu of the summary of results j obtained as part of the required Interlaboratory Comparison Program.
! The proposed change deletes the reference to the EPA program code because it will not be applicable if the NRC changes to an organiza-tion other than the EPA, to do the comparison study. This proposed change is in response to an NRC memorandum from Frank J. Congel, Chief RAB, to Cecil 0. Thomas, Chief SSPB, dated November 4, 1983.
i It does not adversely impact plant safety or change the intent of the present specification. ( (Page 3/4 12-12)
! 7. (TSPS 190), Drinking Water Report Technical Specification Tables j, 3.12.1-2 and 4.12.1-1 A correction of a typographical error on Technical Specification f Tables 3.12.1-2 and 4.12.1-1 is proposed whereby the term "M-3" is i '
. changed to "H-3", which is the correct abbreviation for tritium.
Furthermore, Note a in Table 3.12.1-2 is also revised to provide i additional information concerning the H-3 reporting level for drinking water samples. Note c of Table 4.12.1-1 is revised to clarify that if no drinking water pathway exists within three miles downstream of the site, the LLD of gamma isotopic may be used. Note ; d is also added to Table 4.12.1-1 for clarification of the LLD for ' H-3 in drinking water. These changes are for clarification purposes to indicate that the reporting level and LLD of H-3 may be raised if l no drinking water pathway exists. j i l 168mm11 ,- - l
Attcchu nt / Ridiolegicci A:!=osen nt AECM-84/0338 (6/19/84) Page 12 These proposed changes.do not adversely impact plant safety because they are consistent with NUREG-0473, Rev. 3, Draft 7, with NRC guidance and with present Grand Gulf programs. (Pages 3/4 12-7, 3/4 12-8, 3/4 12-10) - ! 8. (TSPS 191), Dose Rate and Dose from Radioiodines, Technical
' Specifications 3/4.11.2.1, 3/4.11.2.3 and Bases 3/4.11.2.3 i This proposed change substitutes the language " Iodine-131 Iodine-133 Tritium and Radionuclides" for the existing language,
- which is generally of the form "Radioiodines, Radioactive materials l
... and Tritium." This change clearly identifies which specific radioiodines are of concern and is consistent with the Grand Gulf '
ODCM. Additionally, several clarifications to Technical Specifications 3/4.11.2.1 and 3/4.11.2.3 are proposed. Th'e proposed changes correct a typographical error in Surveillance Requirement i 4.11.2.3 and revise several terms and phrases in the same surveillance requirement to make them consistent with the terminology used elsewhere in the technical specifications. ! These proposed changes are consistent with NUREG-0473. Rev. 3 Draft 7, and do not change the philosophy or intent of the technical specifications. Therefore, these changes are considered enhancements which do not adversely impact plant safety. (Pages 3/4 11-8, 3/4 11-13, B 3/4 11-3 and B 3/4 11-4)
- 9. (TSPS 192), CASEOUS RADWASTE TREATMENT (OFFGAS) SYSTEM, Technical 2
Specification 3/4.11.2.4 f The proposed changes to Technical Specification 3/4.11.2.4 are as . follows: 1
- a. Delete the phrase " components as specified in the ODCM" from the LCO to specify that the entire CASEOUS RADWASTE TREATMENT (OFFGAS) SYSTEM shall be in operation.
- b. Revise ACTION a to clarify that the ACTION is required to be performed when gaseous radwaste from tha main condenser air ejector system is being discharged without treatment, thereby providing a quantifiable criterion for entering the subject l ACTION statement.
- c. Add to ACTION a.1 additional reporting requirements to be included in the Special Report.-
- d. Delete the reference to Technical Specification 6.9.1.11 in ACTION b. because this section has been deleted by problem sheet 093 (Special Reporting Requirements).
- e. Revise Surveillance Requirement 4.11.2.4 to be consistent with the revised LCO by verifying that the CASE 0US RADWASTE TREATMENT (OFFCAS) SYSTEM is functioning.
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Page 13 1 l 1 These changes ~ clarify OPERABILITY, surveillance, and reporting tequirements and are consistent with NUREG-0473 Rev. 3, Draft 7, and 10 CFR 20. These changes constitute an enhancement to safety in that they impose additional requirements not presently included in i the technical specifications and conform with recent NRC guidance. (Page 3/4 11-14) l
- 10. (TSPS 194), Operational Radiological Environmental Monitoring Program Technical Specification Table 3.12.1-1 i
A change to " Number of samples and locations" column of Table 3.12.1-1 is requested. For Food Product samples, the change will l require gathering samples of 3 different kinds of broad leaf vegeta-tion, rather than 3 samples of one kind, and will require these samples be taken from two different locations close to the SITE ' BOUNDARY. This proposed change is an enhancement to safety in that j it increases the sampling requirements of the broad leaf vegetation j sampling program. This change is also consistent with NUREG-0473, Rev. 3, Draft 7. (Page 3/4 12-5) I l 11. (TSPS 225), Illegible Figures, Technical Specification Figures i 5.1.1-1, 5.1.2-1, and 5.1.3-1 1 i l Legible copies of the subject figures are submitted to replace the figeres currently found in the technical specifications. All of the subject figures have been redrawn to improve legibility including deletion of typographical lines. These changes are administrative in nature and are submitted in response to NRC concerns over the readability of the subject figures. In addition, Figure 5.1.3-1 has ! been enhanced by adding the effluent release points. (Pages 5-2, l 5-3, 5-4) h l 12. (TSPS 249), Changes to Radiological and Environmental Technical i Specifications
- a. Additional Definition - MEMBERS OF THE PUBLIC, Technical l Specifications 1.24, 3.12.1, 6.9.1.9. 6.15.1.1.e Bases
!. 3/4.11.1, 3/4.11.1.2, Bases 3/4.11.2.1, Bases 3/4.11.2.2, Bases. ! 3/4.11.2.3, Bases 3/4.11.2.7, Bases 3/4.11.4, and Bases i 3/4.12.1 An appropriate definition for the term " MEMBERS OF THE PUBLIC" i has been added to the technical specifications. Present l technical specifications use the term " individual" instead of l M1!MBERS OF THE PUBLIC. The term " individual" can be ! misinterpreted as applying to personnel who work in the plant. This change is being proposed to clarify that MEMBERS OF THE l PUBLIC do not include persons who are occupationally associated with the plant. This change renders the subject technical specifications consistent with .NUREC-0473, Rev. 3. Draft 7 by adding this defined term where applicable. This change does not adversely impact plant safety because the proposed definition is consistent with the NRC guidance and present Grand Gulf programs. (Pages 1-4, 3/4 12-1, 6-18, 6-26. B 3/4 1, B 3/4 11-2, B 3/4 11-3, B 3/4 11-5, and B 3/4 12-1) j
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- b. Additional Definition - UNRESTRICTED AREA, Technical Specifica-e tion:1.46, Bases 3/4.11.1.1, 3/4.11.1.2, Bases 3/4 11.1.4, Bases 3/4 11.2.1, Bases 3/4 11.2.2, Bases 3/4.11.2.3, Technical Specification 6.9.1.9, Techn'ical Specification 6.15.1.e.
1 An appropriate definition for the term " UNRESTRICTED AREA" has been added to the technical specifications to designate those
- areas at or beyond the SITE BOUNDARY for which access is not controlled by the licensee for the purposes of protection of MEMBERS OF THE PUBLIC from exposure to radiation and radio-
' active materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institu-tional and/or recreational purposes. The term UNRESTRICTED
' AREA is added to the subject specifications to be cotisistent with the new proposed definition. This change renders the subject technical specifications consistent with NUREG-0473, Rev. 3, Draft 7 by adding this defined term where applicabir l l This change does not adversely impact plant safety because the proposed definition is consistent with NRC guidance and present Grand Gulf programs (Page 1-8, B 3/4 11-1, B 3/4 11-2, B 3/4 11-3, 6-18, 6-26)
- c. Liquid Effluents - Concentration, Technical Specification '
! 3/4.11.1.1 The proposed revision changes Technical Specification 3.11.1.1 and its associated ACTION statement to specify that only liquid effluents released to UNRESTRICTED AREAS are controlled by the subject specification. The proposed change to Surveillance Requirement 4.11.1.1.2 is a wording change only that adds clarification without changing the requirements of the present i surveillance requirement. These proposed changes are are ; purely administrative and are consistent with NUREG-0473, Rev.
- 3. Draft 7, NRC guidance and present Grand Gulf programs.
(Page 3/4 11-1) i
- d. Radioactive Liquid and Gaseous Waste Sampling and Analysis Programs, Technical Specification Table 4.11.1.1.1-1. Table l1 4.11.2.1.2-1, Bases 3/4.11.1.1, and Bases 3/4.11.2.1.
Proposed changes to the subject tables include the use of microcurie instead of picoeurie, deletion of the "*" footnote
.and deletion of a potentially confusing statement involving
. background determination methods associated with calculating j
the LLD for a radionuclide determined by gamma-ray spectro-metry. The change in units (from pCi to pCi) is a purely administrative change for consistency with published NRC , l guidelines. The deletion of the "*" footnote does not ad- l versely impact. plant safety _because an additional change is proposed to incorporate the deleted information into Bases 3/4.11.1.1 and 3/4.11.2.1. These proposed changes are consis-tent with NUREG-0473, Rev. 3. Draft 7 and are consistent with
~NRC guidance and present Grand Gulf programs. (Pages 3/4 11-3, 3/4 ~ 11-4, 3/4 11-10, and 3/4 11-11. B 3/4 11-1, B 3/4 11-3) i 268mml4 , _. _ _ _ _. , _ , _ _ __
p- , , Attechatnt / Radiological Assessnint AECM-84/0338 (6/19/84) Page 15 i t e. Liquid Effluents - Dose, Technical Specification 3/4.11.1.2 I Proposed changes to Technica,1 Specification 3/4.11.1.2 include appropriate terminology changes from the present use of "an l individual" to "a MEMBER OF THE PUBLIC" and from "from the j _ site" to "to UNRESTRICTED AREA." These changes provide consistency between the subject specification and the new proposed definitions for MEMBER OF THE PUBLIC and UNRESTRICTED i AREA. A change to the associated ACTION statement in also proposed to include a requirement to report, in a Special l Report, those corrective actions that have been taken to reduce the release to within specified limits as well as actions that l l will be taken to ensure that future releases will be in compliance with the specified limits. A "*" footnote is also f' added to specify applicability only if a drinking water supply is taken from the receiving water body within 3 miles downstream of the plant discharge. These proposed changes are l enhancements to safety because they include additional requirements not contained in the current specification and are consistent with NUREG-0473, Rev. 3 Draft 7 as well as NRC guidance and present Grand Gulf programs. (Page 3/4 11-5) l f. Liquid Holding Tanks, Technical Specification 3/4.11.1.4 i A change is proposed to add to ACTION a, of the subject techni-cal specification a requirement to describe, in the next
- Semiannual Radioactive Effluent Release Report, the events which led to entry into the ACTION statement. Reference to Technical Specification 6.9.1.11 is deleted since Section i 6.9.1.11 is deleted by problem sheet 093 (Special Reporting Requirements).- Surveillance Requirement 4.11.1.4 is also changed to clarify that the quantity of radioactive material in "each" of the specified liquid holdup tanks is to be deter-mined. These proposed changes are enhancements to safety in that they impose additional requirements that are' consistent with NUREG-0473, Rev. 3. Draft 7 as well as NRC guidance and present Grand Gulf programs. (Page 3/4 11-7)
- g. Radioactive Effluents Dose - Noble Gases, Technical Specification 3/4.11.2.2, Bases 3/4.11.2.1, 3/4.11.2.2 and Bases 3/4.11.2.3 The proposed change to Technical Specification 3.11.2.2
- clarifies that the air _ dose limits due to noble gas releases from each reactor unit are applicable to those areas at and beyond the SITE BOUNDARY. The change to Surveillance l
Requirement 4.11.2.2 clarifies that dose calculations are made using the contribution from noble gases in accordance with the methodology and parameters-in the ODCM. In addition, a statement is added to Bases 3/4.11.2.2 and 3/4.11.2.3 to ; clarify that effluents are proportioned among the units sharing the Radwaste Treatment System. Bases 3/4.11.2.1 is revised by deleting the statement about effluents from shared units being proportioned among the units sharing the Radwaste Treatment System. This deleted statement does not
- . M - . -
Attechaznt / Rtdiological Asseatmant !
+
- AECM-84/0338 (6/19/84)
Page 16 apply to this specification since dose is dependent on total e efflu.ent release. These proposed revisions are purely administrative changes to maintain consistent terminology throughout the technical specification. They are also j consistent with NUREG-0473 Rev. 3, Draft 7 as well as NRC 1
-guidance and present Grand Gulf programs. (Page 3/4 11-12 J
~
B 3/4 11-3, and B 3/4 11-4) 1
- h. Radiological Environmental Monitoring Program, Technical l
Specification 3/4.12.1 l A revision to ACTION b of the subject technical specification is proposed to require submittal of a Special Report pursuant 1 to Technical Specification 6.9.2 within 30 days instead of the present requirement to submit the Special Report with'in 30 days from the end of the affected calendar quarter. ACTION c is created from the last two sentences of ACTION b to promote understanding and readability. New ACTION c is changed by l specifying broad leaf vegetation instead of fresh leafy I vegetable samples. This change will allow collection of broad leaf vegetation other than edible vegetables to satisfy the requirements of the specification. ACTION c is also modified by deleting reference to Technical Specification 6.9.1 (deleted by problem sheet 093) and requiring instead, that when broad leaf vegetation sampling is relocated, the new location be identified and added to the radiological environmental monitoring program within 30 days. Presently a Special Report is required when samples are unavailable, but this requirement j [j is changed to require reporting the cause(s) of the unavailability of samples and the new locations for obtaining replacement samples in the next Semiannual Radioactive Effluent Release Report. Revised ODCM figure (s) and table reflecting the new locations are also required at this time. A proposed
' change to ACTION c will allow samples that are unavailable to be deleted from the radiological environmental monitoring program as well as the table in the ODCM provided replacement sample locations are identified and documented. Present ACTION c is relettered.to be ACTION d to correspond with the above changes. These proposed changes do not adversely impact plant safety in that they do not reduce the intent of the sampling or-reporting requirements of the current specification. These changes are consistent with NUREG-0473. Rev. 3. Draft 7 as well as NRC guidance and present Grand Gulf programs. (Page
. 3/4 12-1)
- i. Operational Radiological Environmental Monitoring Program Technical Specification Table 3.12.1-1 f
Several changes to Table 3.12.1-1 are proposed for the purpose of rendering the table . consistent with NUREG-0473, Rev. 3. Draft 7. Specific proposed changes are as follows: 268mm16
Attechmsnt / Rrdiological Asssasmant AECM-84/0338 (6/19/84) Page 17
- 1) For AIRBORNE pathways the proposed change requires that three of the'five samples come from locations "close to the SITE BOUNDARY " Also a revision in the terminology for the calculated ground level dose is requested to clarify that the concern is deposition per unit area (D/Q), rather than atmospheric dispersion (X/Q).
- 2) For DIRECT RADIATION pathways or samples, a revision to l the description of the location of the required 40 monitor-ing stations is proposed to clarify that stations do not have to be placed in inaccessible sectors, (e.g. a sector that is unreachable via land vehicles). This change does not adversely impact plant safety because the revised requirement remains consistent with the ODCM. A "*" note is added to specify that accessible sectors are described in the ODCM.
- 3) For Fish and Invertebrate samples, the change deletes the specific requirement to collect " catfish" and replaces it with a more general requirement, "one species of commer-cially or recreationally important fish."
- 4) In the Table Notation several revisions are proposed.
Note a is expanded to provide additional information concerning the number, locations, and collection frequency of samples used in the radiological environmental monitor-ing program. The revised note includes information concerning references, deviation allowances, and reporting requirements, which are consistent with NUREG-0473. Rev. 3, Draft 7. Note e is deleted as it provides information that is appropriate for surveillance procedures rather than technical specifications. Note f is revised to delete reference to Regulatory Guide 4.13 since this is superfluous information not needed in the technical specifications. Notation for Notes f, g, i, and - k are corrected to reflect the deletion of Note e. These changes are enhancements which render the subject table consistent with NUREG-0473. Rev. 3 Draft 7 as well as NRC guidance and present Grand Gulf programs. (Page 3/4 12-3, 3/4 12-4, 3/4 12-5, and 3/4 12-6)
- j. Maximum Values for the Lower Limits of Detection (LLD) Table Notation, Technical Specification Table 4.12.1-1 and Bases
. 3/4.12.1 The proposed change to Note b of the subject Table deletes a statement involving background determination methods associated with calculating the LLD for a radionuclide determined by gamma-ray spectrometry, and deletes the
- i footnote. This deletion does not adversely impact plant safety because the information contained in the
- footnote is moved to Bases 3/4.12.1.
268mm17
Attachaznt / Radiological Asesscssnt AECM-84/0338 (6/19/84) Page 18 A requirement is added to identify and describe in the Annual l e Radiological Environmental Operating Report any circumstances that may render LLDs unachievable. These proposed changes are ; an enhancement to safety, in that they impose additional requirements that are consistent with NUREG-0473 Rev. 3 Draft 7, NRC guidance, and present Grand Gulf programs.(Page 3/4 12-10 and B 3/4 12-1) , k. Land Use Census, Technical Specification 3/4.12.2 I A revision to Specification 3/4.12.2 is proposed to render the specification consistent with NUREG-0473, Rev. 3, Draft 7 . The changes resulting from this revision are in the area of reporting requirements and locations for the land use census. New sample locations are required to be identified in the Semiannual Radioactive Effluent Release Report, rather than the Monthly Operating Report, and the results of the land use i census must now be included in the Annual Radiological Environmental Operating Report. In addition, the references to Licensee Event Report have been deleted in accordance with problem sheet 093 (Special Reporting Requirements). The new distance ( 5 miles) specified for location of the nearest milk animal, the nearest residence, and the nearest garden is an I enhancement to safety in that is ensures more accu 3 ate samples are obtained. Specifying a garden size of 500 ft. provides a
; better definition for sample criteria. (Page 3/4 12-11)
- 1. Interlaboratory Comparison Program Technical Specification
- 3.12.3 A revision to Technical Specification 3.12.3 is proposed to clarify that the analyses for the Interlsboratory Comparison f
Program need only be performed on those radioactive materials that correspond to the samples required to be analyzed by Table 3.12.1-1. The addition of this clarification does not i adversely impact plant safety and is consistent with NRC guidance and present Grand Gulf programs. (Page 3/4 12-12)
- m. Liquid Holdup Tanks, Bases 3/4.11.1.4 ,
i A revision to the subject bases is proposed to clarify that the tanks applicable to Technical Specification 3.11.1.4 are those that are capable of releasing to the environment because they are not surrounded by liners, dikes, or walls capable of holding their contents and do not have overflows or drains that are connected to the liquid radwaste treatment system. This l change will not adversely impact safety and is consistent with ' NUREG-0473 Rev. 3, Draf t 7, NRC guidance and present Grand Gulf programs. (Page B 3/4 11-2) i n. Total Dose, Bases 3/4.11.4 I A revision to the subject bases is proposed to indicate that a Special Report must be submitted whenever plant generated dose l 168mm18
,' ," Attachmsnt / Radiological Assessmant AECM-84/0338 (6/19/84)
Page 19 l exceeds 25 mrem to the total body or any organ, except the e thyroid, which shall be limited to less than or equal to 75 arem, rather than twice the design objective doses of Appendix I. In addition, a .new reference is added 10 CFR 20.405C. .These proposed changes will not adversely impact safety and are consistent with NUREG-0473 Rev. 3. Draft 7 , and NRC guidance. (Page B 3/4 11-5) l
- o. Monitoring Program, Bases 3/4.12.1 I
I A revision to the subject bases is proposed to add a reference
- to 10 CFR Part 50. Specifically, words are added to clarify that the radiological monitoring program discussed in the bases implements Section IV.B.2 of Appendix I to 10 CFR 50.
- This proposed change is consistent with NUREG-0473 Rev. 3, Draf t 7 and NRC guidance. (Page B 3/4 12-1)
- p. Land Use Census, Bases 3/4.12.2 AnadditiontoBases3/4.132isproposedtojustify_the minimum garden size of 50m . Included in this addition are the assumptions used to determine the minimum garden size. This proposed change does not adversely impact plant safety and is consistent with NUREG-0473. Rev. 3, Draft 7 and with NRC i guidance. (Page B 3/4 12-1)
J
- q. Interlaboratory Comparison Program, Bases 3/4.12.3 A revision to Bases 3/4.12.3~is proposed to indicate that the Interlaboratory Comparison Program must be " approved" and that the results of the measurements of radioactive materials in the f environmental sample matrices are " valid for the purposes of I Section IV.B.2 of Appendix I to 10 CFR Part 50." This proposed change is a purely administrative change to clarify the bases l
and is consistent with NUREG-0473 Rev. 3, Draft 7. (Page B 3/4 12-1) r
- r. Annual Radiological Environmental Operating Report,-Technical Specification 6.9.1.7 A revision to Technical Specification 6.9.1.7 is proposed to require deviations from the sampling program identified in I Technical Specification 3.12.1 to be reported in the Annual Radiological Environmental Operating Report. This proposed .
change is consistent with NRC guidance and present Grand Gulf programs and represents an additional requirement not presently in the technical specifications. (Page 6-17)
- s. Semiannual Radioactive Effluent Release Report, Technical Specification 6.9.1.9 i
Revisions to Technical Specification 6.9.1.9 are proposed for the purpose of identifying additional information that should be included in the Semiannual Radioactive Effluent Release 168mm19
Attechnsnt / Radiological Assestsent l
*~
- AECM-84/0338 (6/19/84)
- Page 20 Report; namely, 1) assessments of the radiation doses from radioactive liquid and gaseous effluents received by MEMBERS OF e THE PUBLIC during activities inside the SITE BOUNDARY, along with the assumptions used to make these assessments, 2) causes of any unavailability of samples for the pathway, and 3) the locations for obtaining replacement samples. In addition, the report should include revised figures and tables for the ODCM,
^
which reflect all new sample location (s). The report is also required to include any changes to the ODCM. These proposed changes are enhancements to safety in that they represent additional reporting requirements and are consistent with i NUREG-0473, Rev. 3 Draft 7, NRC guidance and present Grand Gulf programs. (Page 6-18) I t. Monthly Operating Reports, Technical Specification 6.,9.1.10
'- A revision is proposed to delete from Technical Specification ' 6.9.1.10 the requirement to report changes to the OFFSITE DOSE CALCULATION MANUAL and major changes to the radioactive waste treatment systems in the Monthly Operating Report. Such changes will instead be required to be reported in the Semi-annual Radioactive Effluent Release Report. This proposed change will not impact plant safety and is consistent with NUREG-0473, Rev. 3. Draf t 7 and NRC guidance. (Page 6-19)
- u. Offsite Dose Calculation Manual (ODCM) Reporting Requirements, Technical Specification 6.14.2.1 A revision to Technical Specification 6.14.2.1 is proposed to indicate that changes to the ODCM shall be reported in the Semiannual Radioactive Effluent Release Report rather.than in the Monthly Operating Report. This proposed change will not adversely impact plant safety and is consistent with NUREG-0473, Pev. 3, Draf t 7 and NRC guidance. (Page 6-25)
W i 168mm20
*
- Attechment / Radiolcgical Assaassent AECM-84/0338 (6/19/84)
Page 21 E . 6 E. PROPOSED TECHNICAL SPECIFICATION CHANGES (AFFECTED PAGES ARE PROVIDED IN THE ORDER OF ASCENDING PAGE NUMBERS.) 4 I a O r 4 268mm21
w _. . . - - --. - l ( - DEFINITIONS l
- E-AVERAGEDISiNTEGRATI'ONENERGY 1.11 I shall be.the average, weighted'in pro ortion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sv.n of the average beta and gamma energies per disintegration, in MeV, for isotopes, with. half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.
EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME l 1.12 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time - interval from when the monitored parameter exceeds its ECCS actuation setpoint l ! at the channel sensor until the ECCS equipment is capable of performing its , safety function, i.e., the valves travel to their required positions, pump i discharge pressures reach their required values, etc. Times shall include ! diesel generator starting and sequence loading delays where applicable. The , response time may be measured by any series of sequential, overlapping or ; total steps such that the entire response time is measured. , END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME
- 1.13 The END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval to complete suppressicn of the electric arc between the fully open contacts of the recirculation pump circuit breaker from initial movement of the associated:
- a. Turbine step valves, ari
- b. Turbine control valves.
The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured. FRACTION OF LIMITING POWER DENSITY 1.14 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR existing at a given location divided by the limiting LHGR for that bundle type. FT./.CTION OF RATED THERMAL POWER 1.15 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL POWER di/ided by the RATED THERMAll POWER; . I FREQUENCY NOTATION 1.16 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1. GASEOUS RADWASTE TREATMENT (OFFGAS) SYSTEM 1.17 The GASEOUS M installedtoreduceT._.gATMENT(OFFGAS)SYSTEMisthesystemdesignedand
,a gaseous effluents by collecting primary coclant
$) ,
system offgases from the primary system and providing for delay or holdup for the l purpose of reducing the total radioactivity prior to release to the environment. l ( l GRAND GULF-UNIT 1 1-3 _
DEFINITIONS IDENTIFIED LEAKAGE - 1.18 IDENTIFIED LEAKAGE shall be:
- a. Leakage into collection systems,.such as pump seal or valve packing leaks,.that is captured and conducted to a sump or collecting tank, or
- b. Leakage into the drywell atmosphere from sources that are both l
.specifically located and known either not. to interfere with the opera-
, tion of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE. ISOLATION SYSTEM RESPONSE TIME . .
~
1.19 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at.the i channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured. LIMITING CONTROL ROD PATTERN , 1.20 A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core being on a tharmal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPF.. LINEAR HEAT GENERATION RATE 1.21 LINEAR HEAT GENERATION RATE (L5iGR) shall be the heat generation per unit length of fuel rod. It is the integral of the heat flux over the heat ,' transfer area associated with the unit length. LOGIC SYSTEM FUNCTIONAL TEST j 1.22 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components,
- 1.e., all relays and contacts, all trip units, solid state logic elements, i
etc., of a logic circuit, from sensor through and including the actuated
<f avi ca . to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.
MAXIMUM FRACTION OF LIMITING POWER DENSITY
! 1.23 The MAXIMUM FRACTION OF LIMITING. POWER DENSITY (MFLPD) shall be the highest value of the FLP0 which exists in the core.
IM EMT -* ; 4 MINIMUM CRITICAL POWER RATIO e4
- 1. 2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.
GRAND GULF-UNIT 1 1-4 AA4,a ya. r
. _ . . _ _ , , _ _ _ _ , , ._ _ _ - _ , _ _ . - _ . - . , . , , - . __,__,_r__-,,
4-i l i INSERT TO DEFINITIONS, PAGE 1-4 MEMBER (S) 0F THE PUBLIC 1.24 HEMBER(S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors or vendors. Also er.cluded 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 recreational, occupational or other purposes not associated with the plant. l l l 268abm28
DEFINITIONS OFFSITE DOSE CALCULATION MANUAL.(00CM)
- 1. The OkFSITE DOSE CALCULATION MANUAL s~ hall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous lh and liquid effluents and in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints. It shall also contain a table and figure defining current radiological environmental monitoring sample locations.
OPERABLE - OPERABILITY 1.2 A system, subsystem, train, component or device shall be OPERABLE or have OP. ABILITY when it is capable of performing its specified function (s) and ldE. 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). OPERATIONALC5NDITION-CONDITION s 1.7fAnOPERATIONALCONDITION,i.e., CONDITION,shallbeanyoneinclusive E combination of mode switch posiU ;n and average reactor coolant temperature as d specified in Table 1.2. PHYSICS TESTS
't 1.25 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and ll
- 1) described in Chapter 14'of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.
PRESSURE BOUNDARY LEAKAGE _ 1.NPRESSUREBOUNDARYLEAKAGEshallbeleakagethroughanon-isolablefault lE in a reactor coolant system component body, pipe wall or vessel wall. E' GRAND GULF-UNIT 1 1-5 Awb=T #8 -
DEFINITIONS ( PRIMARY CONTAINMENT INTEGRITY ,, t . 1.37 PRIMARY CONTAINMENT INTEGRITY shall exist when: h
- a. All containment penetrations required to be closed during accident conditions are either:
- 1. Capable of being closed by an OPERABLE containment automatic isolation system, or
- 2. Closed by at'least one ma'nual' valve, b5'ind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.4-1 of Specification, 3.6.4.
- b. The containment equipment hatch is closed and sealed.
- c. Each containment air lock is OPERABLE pursuant to Specification 3.6.1.3.
- d. The containment leakage rates are within the limits of Specification 3.6.1.2.
- e. The suppression pool is OPERABLE pursuant to Specification 3.6.3.1.
- f. The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows or 0 rings, is OPERABLE.
PROCESS CONTROL PROGRAM (PCP) 1r3-1 The PROGE-SG-GONTROL PROGRAM-shah-conte 4*-tt: ::mpling, analy:is, :nd fer ulation determination by which SOLIDIFICATION cf r:dic::tive w::te: fr:: y 1iquid :ystem; i: :::; red. ru .=
- o PURGE - PURGING -
.L.# PURGE or PURGING is the controlled process of dischargi,ng air or gas from aconfinementtomaintaintemperature, pressure, humidity,concentrationorother.g <
operating condition, in such a manner that replacement air or gas is required to purify the confinement. I RATED THERMAL POWER
- s 1 1.31 RATED THERMAL POWER shall be a total reactor core heat transfer rate to 4 the reactor coolant of 3833 MWT. ,
N l l A~rmeer N o. GRAND GULF-UNIT 1 1-6 , - 1
l INSERT TO DEFINITIONS, 1.31 PAGE 1-6 1.32 The PROCESS CONTROL PROGRAH rhall contain the current formula, sampling, analyses, tests, 2nd determinations to be made to ensure that the processing 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 compliance,with 10 CFR Part 20, 10 CFR Part 61, 10 CFR Part 71, and Federal and State regulations, burial ground requirements and other requirements governing the disposal of the radioactive waste. 268abm23
i DEFINITIONS l { REACTOR PROTECTION SYSTEM RESPONSE TIME 1.3f REACTOR PROTECTkON SYSTEM RESPONSE TIME shall be the l$" tim when the monitored parameter exceeds its trip set;)oint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by any series of segeantial, overlapping or total steps such that the entire response time is meas'. ired. REPORTABLE OCCURRENCE 6 1.35 A REPORTABLE OCCURRENCE shall be any of those conditions specified in Spepifications 6.9.1.12'and 6.9.1.13. -- h" d ROD DENSITY ,
'7 n o-1 5 ROD DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted lg is equivalent to 100% R00' DENSITY.
SECONDARY CONTAINMENT INTEGRITY 1.3f SECONDARY CONTAINMENT INTEGRITY shall exist when: l r4
- a. All Auxiliary Building and Enclosure Building penetrations required to be closed during accident conditions are either:
- 1. Capable of being closed by an OPERABLE secondary containment i automatic isolation system, or
- 2. Closed by at least one manual valve, blind flange, or deactivated automatic valve or damper, as applicable, secured in its closed positien, except as provided in Table 3.6.6.2-1 of Specification 3.6.6.2.
- b. All Auxiliary Building and Enclosure Building equipment hatches and blowout panels are closed and sealed.
- c. The standby gas treatment system is OPERABLE pursuant to Specification 3.6.6.3. ,
- d. The door in each access to the Auxiliary Building and Enc 1'sure o
Building is closed, except for normal entry and exit.,
- e. The sealing mechanism associated with each Auxiliary Building and Enclosure Building penetration, e.g. , welds, bellows or 0-rings, is
. OPERABLE.
3 . GRAND GULF-UNIT 1 1-7 A dmenT M* -
DEFINITIONS J l SHUTDOWN MARGIN 1.3f SHUTDOWN MARGIN shall be the subcritical' or would be subcritical assuming all control rods are fully m a' ount of reactivity by which the rom: tor is' hN inserted except for the single control rod..of highest reactivity worth which is assumed to be fully withdrawn and~the reactor is in the shutdown condition; cold, i.e. 68'F; and xenon free. An4r* -
. p' a ' SOLIDIFICATION
- 1. N SOLIDIFICATION shall be the a s c,monversion
. . .a . .. .e ____3.u_ of -- N: ti : wastes 2 ' m 'i u!d -l$d g:._...u._______... i m_4 ,_ _ ___u ,1-i .selid with def4afte uelu e end s? ape, beunded by e st 9!e rurfece ef dictiret > "
i
'eutl4ae aa e!! Sides (# ee-stead 4aa)- Wie a. fam ht- mW3 sW / h **J #
b.e . d v. wad re pt. e m ent,s, SOURCE CHECK ' e 1.4IASOURCECHECKshallbethequalitativeessessmentofchannelresponse $ when the channel sensor is exposed to a radioactive source. STAGGERED TEST BASIS 1.4} A STAGGERED TEST BASIS shall consist of: l$w ,
- a. A test schedule for n systems, subsystems, trains or other-I desig'nated ccmponents obtained by dividing the specified test
( interval into n equal subintervals. I b. The testing of one system, subsystem, train or other designated - component at the beginning of each subinterval. THERMAL POWER 1.41 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. h UNIDENTIFIED LEAXAGE 1.d UNIDENTIFIED LEAXAGE shall be all leakage which is not IDENTIFIED LEAXAGE.
- ZNSERT -* UNMastnuYab AREA la6 ' p e.
8 VENTILATION EXHAUST TREATMENT SYSTEM 1.4fAVENTILATIONEXHAUSTTREATMENTSYSTEMisanysystemdesignedandinstalled t.c., reuuce gaseous radiciodine or radioactive material in particulate form in I effluents by passing ventilation or vent exhaust gases through charecal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates 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 Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components. VENTING ,
,1.4EVENTINGisthecontrolledprocessofdischargingairorgasfroma confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in. system names, does not imply a VENTING process.
- GRAND GULF-UNIT 1 1-8 Aw &,eNo._
- - - - , . . . - , - . . . . , . _ , ,- . . _ , , , . . , , . , . . , . . , + _ . . , - . - - , , , - - , . - . , . + . .-nw,.e- ,..e-.,, , . . , - , ,n, ,n-,,
n . a INSERT A TO DEFINITIONS Page 1-8 SITE BOUNDARY 1.40 The SITE BOUNDARY shall be that line beyond which the land is 'neither owned, nor leased, nor otherwise controlled by the licensee. ~ INSERT B TO DEFINITIONS, Page 1-8 UNRESTRICTED AREA 1.46 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 MEMBERS OF THE PUBLIC from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for f.ndustrial, commercial, institutional, and/or recreational purposes. 4
.268mm23
3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION i . LIMITING CONDITION FOR OPERATION unAssrnscTKb hAEAG ', lye ) JTiver.Ts
~
3.11.1.1 The concentration of radioactive material release fr = th: :it: to .
; = r= trict:d : = = (see Figure 5.1.3-1) shall be limited to the concentrations >i i
specified in 10 CFR Part 20, Appendix 8. Table II, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained " nobl# gases, the concentration shall be limited to 2 x 10 4 microcuries/ml total activity. APPLICABILITY: At all times. in l' a:)
- W Iu
- N s to ACTION:- . uNA5YrMrc.7En AAs'es With the concentration of radioactive material released c= th :it: exceeding I the above limits, imediately restore the concentration to within the above N limits.
SURVEILLANCE REQUIREMENTS 1 . 4.11.1.1.1 The radioactivity content of each batch of radioactive liquid waste shall be determined prior to release by sampling and analysis in accord-ance with Table 4.11.1.1.1-1. The results of pre-release analyses shall be used with the calculational methods in the ODCM to assure that the concentration at the point of release is maintained within the limits of Specification 3.11.1.1. 4.11.1.1.2 Post-release analyses of samples composited from batch releases
, shall be performed in accordance with Table 4.11.1.1.1-1. The results of the g p=i=: ;=t r;1== :=1y=: e:11 5: =d eie th =?=1:ti:--! =tb
- i. tu com t: === tut th: ===t=ti:= :t te ;: int :f =? == . =
l =inui=d within tu 11:it: : f : p u i f f = t i n :. n . :. 1 .
% at ~ s :ry . ,..I fat, Jn I,, usa) ;,. . L,. e . w;rs, a,- ,.
l stA*M*sy *.J m =~ cars io, ti,, com , a uena y j>that the codenn% tier.s # 1% a. plnr ,F ratsa.1, art.
& ;n ta.'e s ) ,wlrA'.o. rh a. I.'~;rs e .F sp ;Fiut;. s. st.l.l.
GRAND GULF-UNIT 1 3/4 11-1 Amendment No. g l l I l
- TABLE 4.11.1.1.1-1 (Continued)
RADI0 ACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM TA'BLE NOTATION
- a. The LLD'is the smallest concentration of radioactive material in a sample i .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. 1
~
I For a particular measurement system (which may include radiochemical, separation): .
- 4.66 s b ::
LLD = E - V - 2.22. *Y
~
exp(-Aat) - e-v Ex tD d where LLD ish.he"apriori h lower limit of detection as defined above (as p/Ciperunitmassorvolume). (Current literature defines the lEd LLD as the detection capability for the instrumentation only, and the MDC, minimum detectable concentration, as the detection capability for a given instrument,,p.rocedure, and type of sample.) s is the standard deviation of the background counting rate or of b the counting rate of a blank sample as appropriate (as counts per minute) E is the counting efficiency (as counts per disintegration) V is the sample size (in units of mass or volume) x #*' miu wri,e g 2.223 is the number of disintegrations per minute per 7 ::;-. u g Y is the fractional radiochemical yield (when applicable) A is the radioactive decay constant for the particular radionuclide At is the elasped time between sample collection (or end of the sample collection period) and time of counting The value of su used in the calculation of the LLD for a particular measurement system should be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicated variance. v__,....,__m. ,,n ,._ . __2,__...,,2. 2_ . . __ , . . a u . , ..__....m
!._ i Zl'Z "*.2"'u 7_i:l._' 2:!. U iliC 2:';E ..Z7.3 L_._n.3 ,___ ., '
5 EEEk N5Ih$0:5iE ElEE555 hSEESt 'T55E ESMIE[kypicalla5ues~of E, V, Y and at should be used in the calculation. It should be recognized that the LLD is defined as an a priori (before f the fact) limit representing the capability of a measurement system and ( e-not as a, costeriori (after the fact) limit for a particular measurement.# ls$ 3/4 11-3 4 %, b ,# #p. _ l GRAND GULF-UNIT 1 r _
. . l l
TABLE 4.11.1.1.1-1 (Continued) RADIOACTIVE LIOUID WASTE SAMPLING AND ANALYSIS PROGRAM TABLENOTATION(Continued) l
- b. 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 which is representative l of the liquids released.
- c. 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 to assure representative sampling. .
- d. The principal gamma emitters for which the LLD specification ap lies exclusively are the following radionuclides: Mn-54, Fe-59, Co .,8, Co-50, Zn-65, Ho-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and recorted. Other ;
peaks which are measurable and identifiable, together w'ith the above nuclides, shall also be identified and reported.
~
mv . . wmp i n. J :,w.. v.. vi t6 i.LO, .nd vt.'.. J.L. d:vn limit., . the fo - (1) HASL Procedur HASL-300(revisedann . (2) Currie L. A. , " Limits
~
a lon and Quantitative , Determination-Applicat 1 . " Anal. Chem. 40, p 586-93 (1968 - ' N (3) Har - , . . " Detection Limits for Radioisotopic Cou nhiques, inu ,- o < ,-sidi a w2 nin, a renn s ny o nn,.+ aow-9u7 t 1,m. 99 107 GRAND GULF-UNIT 1 , 3/4 11-4 A nJae # #e. l
~
RADI0 ACTIVE EFFLUENTS i DOSE 2 I E I I LIMITING CONDITION FOR OPERATION s n pGMBEM OF THG 9*dDLEC. e-3.11.1.2 The dose or dose commitment tqA:n 'nd hidni from radioactive materials ' in liquid effluents released, from each reactor unit 'r th it Assas, j Figure 5.1.3-1) shall be limited: Ir5 u:u:nasr:ar:crr:a(see
- a. During any calendar quarter to less than or equal to 1.5 mres to the ~
total body and to less than or equal to 5 arem to any organ, and
- b. During any calendar year to less than or equal to 3 mrem to,the total body and to less than or equal to 10 arem to any organ, i APPLICABILITY: At all times. .
g,g y, y g, , 4,, 4
** ""J O * ""#*' " * "" # # l l ACTION- l i
- a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, in lieu of l any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specifica-tion 6.9.2, a Special Report which identifies the cause(s) for ,
exceeding the limit (s) and defines the corrective actions to e e-taken to ensure that future releases will be in compliance with
- d Onh:th: 2.11.1. 2. ras abava. I;a in . 4
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
4 l SURVEILLANCE REQUIREMENTS 4.11.1.2 Dose Calculations. Cumulative dose contributions from liquid
.Tiiuerits s E 31 days fc(hall f., tA.
be urr t la,J.r. determined .in 7A,3,1,fy aJ accordance fuwith rers 0 the 4 9.e.rrar a.iJ rAa c.wrear of th , W I.nJar ye.r i i I Thi" QacialRaynorr skall mise ina.l.,Ja u) rh re s.olts of \ r*dI*l*s***l *nalpas J ri,a. s,sorea tha vnd,'elo;larl l y t ,, Jeta y;,,;,x,n,',,,) y,y; w a ra wyax,) r in) h Wel;'> w M. re ) u n , , ray,,;,a n a,,r, a I , 40 CF/t Part 141,4, I I j f
*Ayy1,'ea L la. enI es*yly Ie tu ha e, .9een , ri,s reulve'ns $e a ar s :L, if Jr;p).kln's.
>- , wear.r,-- ,1 s , ,i. , s;sa.. , .
GRAND GULF-UNIT 1 3/4 11-5 Amendment No. 8, l, 1
RADI0 ACTIVE EFFLUENTS LIQUID WASTE TREATMENT l i I i f LIMITING CONDITION FOR OPERATION ' 3.11.1.3 The liquid radwaste system :: ;::: t: = :;::i'i:d f- th: C:'C" :5:1' lO86 I 5: 0"!R?"LE. 'he apprepri te perti:n ef the :ytte shall be used to reduce i th2' radioactive materials in liquid wastes prior to their discharge when the j due to the liquid effluent 'r= th: :fte (see
- ?:tf>-projecteddosfyperf:dwouldexceed0.06mremtothe.totalbody 086 Figure 5.1.3-1) ' : 31 -
or 0.2 mrem to any organjin a 31 day period. , Pom each reactor un t APPLICABILITY: At all times. to UNRESTRICTED AREAS ACTION:
- a. With the 'fquid r:de::t: treat nt :y:t: '9:;;r:51: f r ter th:- l086 I
31 d:y: = with radioactive liquid waste being discharged without I treatment and in excess of the above limits, in lieu of any other report required by Specification 6.9.1. prepare and subtait to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report which includes the following information: Id=ti f f =tf = cf .th: !=;;r:b!: :q f; : t : : 5:ytt:= :nd 1. th: 70:007 f:P # 9:F:P:bity. l056 hsett i 2. Action (s) taken to restore the inoperable equipment to OPERABLE l status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
- b. The provisions of Specifications 3.0.33 3.0.4 :nd 5.".1.11 are not lO86 applicable. and
...-..-. . . . ..*" '1 EQUI REMENTS I
l ( freM each reacter un'it . .A ' UNRESTRICTED AREAS i 4.11.1.3.1 Doses due to liquid releases to = r::trf:t:d = ::: shall be 086 1 l projected at least once per 31 days, in accordance with the 00CM. . I" l 4.11.1.3.2 T Th: l';;id red;=t: :y:t:= :=;==t:mathe :;;:if':*Iof_f' A#d **"*-t.f:00C":h:1' 5 d = =:tr:t:d 0"ER.^"LE by :; = :tir.; th: l';;'d 7:d = t: tra ta=0 :y:t:r 086
- f; =t f=
- t 1
- =t :e !.,2:= :: 1:=t =:: ;= e: say: ='::5-the '!; !
nent: :y:::: ha ::= ettitue t: ; ==: nef =:t';; if;;ie: d='r; th: n.., _m,.... The installed li 9uid redweste system sbll be demonsteated CPERAELE by meeting Specifications S.ll.l.l and 3.lI.l.2 . , e GRAND GULF-UNIT 1 3/4 11-6 Amendment No. 7% l , e l
a 99 ! INSERT TO ACTION a.1 0F SPECIFICATION 3.11.1.3, PAGE 3/4 11-6 Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems which resulted in liquid radwaste being discharged without treatment, and the reason for the
- inoperability, l
t i l a l 4 i e : I
)
268aba24 l
RADIOACTIVE EFFLUENTS l LIOUID HOLOUP TANKS ' LIMITING CONDITION FOR OPERATION . 3.11.1.4 The quantity of radioactive material contained in any outside tempo . rary tank, not including liners for shipping radwaste, shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases. APPLICABILITY: At al1 times. ACTION: i
- a. ' With the quantity of radioactive material in any of the above specified tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tanks and within 48 hours reducethetankcontentstowithinthelimity b.
The provisions of Specifications 3.0.3}3.0.4=: 5.0.1.11 are net applicable. y SURVEILLANCE REOUIREMENTS quc k dF 4.11.1.4 Thequantityofradioactivematerialcontainedihtheabove h specified tanks shall be determined to be within the above limit by analyzing N a representative sample of the tank's contents at least once per 7 days when radioactive materials are being added to the tank. duer:ba. ske a n.,rs lu)],,9 ro rka udltl*n In L(andrk' . v vt C.~;ueval ot. h ,arlrs E W r $ < % s. M+rr. O GRAND GULF-UNIT 1 3/4 11-7 Amedmod #8. l
~
-S . s ~ (J s f
RADI0 ACTIVE EFFLUENTS 3/4.11.2 GASEOUS EFFLUENTS . e DOSE RATE . LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site 4(see Figure 5.1.3-1) shall be limited to the following: do acess at ya 4 eye 4 +A= Es?A Sou 4)C44 y
- a. For noble gases: Less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin, and
- b. ForabrS$ciY$4Ytritiumandall-$4 e[tE I+=^1e in l5 particulate form with half lives greater than 8 days: Less than or equal to 1500 mrem /yr to any organ.
APPLICABILITY: At all times. ACTION: With the dose rate exceeding the above limits, immediately decrease the release rate to within the above limit (s). I i SUREVEILLANCE REOUIREMENTS 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 ::thed: and pr*: &re of the ODCM. mWe' del.9 y P* '* " ' 4* ' 8 yeJ,'w.-13/, J. Jim.-133 r .),'.% lJas1 4.11.2.1.2 The dose rate due to a sdie4ed'- :, tritium and top :dferrt91: r:t:rf & in particulate form with half lives greater than 8 days, eth:r th:2
. .'.' sen, in gaseous effluents shall be determined to be within the above f limits in accordance with the ethet --d prece& :: of the ODCM by obtaining representative samples and performing analyses in accordance with the sampling .
and analysis program specified in Table 4.11.2.1.2-1. sn a fh o d efe n a s d p a r a m e + e r g
. O GRAND GULF-UNIT 1 3/4 11-8 A n ,, g,, , ,3,, ,_,,
l
. TABLE 4.11.2.1.2-1 (Continued)
RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM
- TABLE NOTATION a, The LLO is the smallest concentration.of radioactive materjal in a sample that will yield a net count (above system background) that will be detected j 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 ine'lude radiochemical separation):
. 4.66 s b "
2.22
- LLO = E - V Y *
**P C ~Mt) 7X80a ,
f where LLD is the "a priori" lower limit of detection as defined above (as
~p #Ci per unit mass or volume). (Current literature defines the %'d LLD as the detection capability for the instrumentation only, and the MOC, minimum detectable concentration, as the detection capability for a given instrument, procedure, and type of sample.)
s is the standard deviation of the background counting rate or of b the counting rate of a blank sample as appropriate (as counts per minute) E is the counting efficiency (as counts per disintegration) V is the sample size (in units of mass or volume) micecerlet e 2.223sisIO*the number of disintegrations per minute per ; m er ': r
#4 Y is the fractional radiochemical yield (when applicable)
A is the radioactive deca'y constant for the particular radionuclide ! at is the elasped time between sample collection (or end of the sample collection period) and time of counting The value of s used in the calculation of the LLO for a parti'cular . measurementsyItemshouldbebasedontheactualobservedvarianceofthe background counting rate or of the counting rate of the blank samples (as l appropriate) rather than on an unverified theoretically predicated
. variance.
i s. ..,,,,,. <.. su ,,n u. . ..au.....,sa. a. ...s..a uu .......u
- __ H _z_;_., M _ i;n ._z_] X Eij iiis.'. r ;;_ . 11.3 E _. 1 3 ,___ ., ' '
e l EEEEkh$5'[:N:fi5:ENE"--E55hhEE5Et'-EEE'EUf$E TypIcallaluesoI $ 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 a, posteriori (af ter the fact) limit for a particular measurement.I $ t GRAND GULF-UNIT 1 3/4 11-10 AmanManf M8. l l
cr . . . .. . . . - . . . l f TABLE 4.11.2.1.2-1 (Continued) l NADIDACTI'VEGASE0b5WASTESAMPLINGANDANALYSISPROGRAM TABLE NOTATION (Continued)
- b. Analyses shall also be performed following startup from cold shutdown, or a THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period.
- c. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours after changing or after removal from sampler.
- Sampling and analyses shall also be performed at least once per 24 hours for at least 7 days following each shutdown, startup or THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER in one hour. When samples collected for 24 hours are analyzed, the corresponding LLD's may be increased by a factor of 10.
- u. 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 Specifications 3.11.2.1 and 3.11.2.3.
- e. The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-C8, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for
. particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, sh,all also be identified and reported.
are complete discussion of the LLD, and other detection limits, the foi o (1) HASL Procedure HASL-300 (revised annu . i (2) Currie, L. A. , " Limits itativ on and Quantitative h Determination - Application u. try" Anal. Chem. 40, 4 586-93 (1968). hl (3) Hartwell ., Detection Limits for Radioisotopic Tecnhiques," , c Richfield Hanford Company Report ARH-2537 (June 22,. l I L ' 1 . l E 1 GRAND GULF-UNIT 1 3/4 11-11 Am,.,Weat No. -
~_
,l l
1 RADI0 ACTIVE EFFLUENTS l D0SE - N0BLE GASES . LIMITING CONDITION FOR OPERATION 3.11.2.2 The air' dose due to noble gases released in gaseous effluents, from t each reactor unit, from the sit (see Figure 5.1.3-1) shall be limited to the le. N following: arms er .J A, y. . .) cA, ssrs nowanny During any calendar quarter: Less than or equal to 5 mrad for gama
- a. ..
radiation and less than or equal to 10 mrad for beta radiation, and During any calendar year: Less than or equal to 10 mrad for gama b. radiation and less than or equal to 20 stad for beta radiation. APPLICABILITY: At all times. ACTION:
- a. With the calculated air dose from the radioactive noble gases in gaseous affluents exceeding any of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Comission within 30 days, pursuant to Specification 6.9.2, s.
Special Report which identifies the cause(s) for exceeding the limit (s) , and defines the corrective actions to be taken to ensure that future
- releases will be in compliance with Specification 3.11.2.2.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS for nebla *)ases Cumulative dose contributions ' :: :::rre 4.11.2.2 Dose Calculations.*"
- for the current calendar quarter and current calendar ;- y deterstned in accordance with the CDCM at least once per 31 days.
k atkjeley, e,J p arers in 1(* y l ) l Amendment No. g. 3/4 11-12 ! GRAND GULF-UNIT 1
~
RADICACTIVE EFFLUENTS 2*Ded ss#,reein 833,vestrum aasa Ratroce suess DOSE - ' IN PARTICULATE FORM /b T"IT!"" g," 3 LIMITING CONDITION FOR OPERATION s l>En\ SEA O C M PuSLK. 11, ~ 4-d!"!t#from trit?=,ye.s3 hdhn :nd ,7:dh::t.=totta< .i n +,. w p.4 7:dS T
, g g __.._.: f.The dose to r) in particulate form with half-lives greater than 8 days in gaseous i y _fluents of released, from each reactor unit, from the sitg (see Figure 5.1.3-1) j,g i shall be limited to the following: /
i 4e e.rees e& sna heye*L 44e. SITA Douabofot y
- a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ, and ,
- b. During any calendar year: Less than or equal to 15 area to any organ.
APPLICABII.ITY: At all times. gddre, I ACTION-' j ! a. With the calculated3e/ doseAfrom c- 831 , T.J.*a the release e - r/.33, of-triti ,"r:d hfe/ hd G h ns, an d rad t.el/
- er r
- df r tive --te-SM in ,.:thcht: form, with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specifica-2 tion 6.9.2, a Special Report which identifies the cause(s) for j exceeding the limit and defines the corrective actions be taken l 'e_,
i to ensure that future releases will be in compliance w th Specification 3.11.2.3.,
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
I f4=+ Aa s a & e.a 4.r 4. ,. g.., f. gg ,,, f SURVEILLANCE REQUIREMENTS
#'*f*M *"#/"" ** */M #
Je&/w- #31,2 4 me. . ill . +'i+/=m **A **d/**a.*>,3es _ l 4.11.2.3 Dose Calculations. Cumulative dose contributions from tit'=, A E j
- dhhdhn, =d r: din:ti= ntrht in particulate form alf-lives greater than 8 days for the current calendar quarter and curren calendar year
{ ODCM at least once pe ;31 days. p' shallbedeterminedinaccordancewiththe/ %g s s. t k . 4 . 1e m u 6 p a r - e.4 a r s i s n e. l GRAND GULF-UNIT 1 3/4 11-13 Amendment No. 8,__ l
RADI0 ACTIVE EFFLUENTS
- GASEOUS RADWASTE TREATMENT
~
LIMITING CONDITION FOR OPERATION . 3.11.2.4 The. GASEOUS RADWASTE TREATMENT (OFFGAS) SYSTEM :::;:n:nt: x :::c'" :d ggg in th: ::" shall be in operation. APPLICABILITY: Whenever the main condenser air ejector system is in cperation. ACTION ~- .I "8
- system being dischaeged' without treatment for i
- a. With,the CAS: U: P,AC'.lAST: TT.:AT" NT (CIPCAS) SYST ?: in:p;r::' 2 : f:r more than 7 consecutive days, in lieu of any other report required l192 by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information:
l Explanation of why 3aseous rodwaste was being discha.rged without treatment, lj ildentification of the inoperable equipment or subsystems and l192 the reason for inoperability,
- 2. Action (s).taken to restore the inoperable equipment o OPERABLE i status, and
- 3. Summary description of action (s) taken to prevent a recurrence.
and
- b. The provisi,ons of Specifications 3.0.3 A 3 .0.4 :nd 5.9.1.11 are not ll97 applicable.
C. which resulted in gaseous radwaste SURVEILLANCE REOUIREMENTS be'eng discha.rged without treatmentj 4.11.2.4 The CASECUS RAO!AST TREAT"ENT (OTFCAS) SYSTE!-! :::p:n:nt: :p::i'f:d
'- th OCC 05:11 5: d::en:tr:ted OI A"LC by :per:tir.g th: CAS:005 ",AC'.",STE
; > =? %.m SYSTE?' :: p:n:nt: :: d:::ribed in th: 00C!! f r :t ?:::t 20 e' ete: N
-at 1:::t : :: per 92 d:y: erle:: th: :y-t = 5:: 5::r utt'fzed te p-ecess
! r:dfe::tfv: ;:: dur'n; th: preef:' : 92 d:y:. . The insivamewis speddled in Me CDcht shall 6e checked every 12 hours whenever h main condenser a'm ejector. sys(.am is in operafion 14 ensure that ne GAStous RAowAsrF TREATMENT l (oFFGks) SYSTEM is funcUening. 1 l l l i GRAND GULF-UNIT 1 3/4 11-14 AmendmeWI No. l
RADIOACTIVE EFFLUENTS VENTILATION EXHAUST TREATMENT i~ t 3 LIMITING CONDITION FOR OPERATION 3.11.2.5 The hppr:pri:t: p:rti:n: ef th: VENTILATION EXHAUST TREATMENT SYSTEM shall be OPE"ABLE and be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected cumul::fv: dose due to gaseous 086 effluent releases from-th: :ite (see Figure 5.1.3-1) in a 31 day period would lg o k each reactor un.t exceed 0.3 mrem to any organ. i to areas at and APPLICABILITY: At all times
- beyond the SITE BOUNDARY ,
l086 ACTION: Other than when he VENTILATioM EXWAUST
- a. With th:TREATMENT VENT!LAT!0F EX"^.UST system is undergYSTE" "E?T"ENT . in:pfor r:b1:oing
- r: routine ma 086 th:r 31 d:y:, Or with gaseous waste being discharged without treat.
ment and in excess of the above limits, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report
- which includes the following information:
1 Identi'te: tier Of the ineper:b1: : quip::nt er : b:y:t::: :nd ggg Ihsert /. th 700 0F ICT OE070bilit2'
- 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 Specifications 3.0.33 3.0.4 :nd E.9.1.11 are not lO86 applicable. pg enouevii w e DEOUIREMENTS g ea.ch teat. tar nwit in A.ress at and beyond the SITE SouNDARY 4.11.2.5.1 Dosesduetogaseousreleasesfromith :it: ,shall be projected at hlO86 leastonceper31daysinaccordancewith[.the00CM.YENTIL^.T 4.11.2.5.2g S: metb 1e p a.n 4
. . :h:11 5: d:::::tr:ted OPER^.SLE by :p:r: ting th: YENTIL^ TION EX"A"ST TRE^.T"E"T SYSTE". :;;f;::nt f:r 086
- t le::t 30 -imet::, :t 1:::t :n:: p;r 92 d:y: unit:: th: :ppr:prict: :ytt:-
-5:: i;;n utili::d t: pr :::: r:di ::th; ;;;;;;;; :f flu;nt: during th: pr:viru:
92 dry:. I j a Not applicable to Turbine Building ventilation exhaust unless filtration media is installed The insiellet VENTILATioM nWAutT TREATMENT S YSTEM shall be. dernenstrated oPERABt.E.' hy m e eting Specihcatsens 3.ll.2.l 7 GRAND GULF-UNIT 1 3/4 11-15 - Amendment Mo. - l l _ _n.md__S. It . 2. 2. or 3. ll . Z. t .__ . _ _ _ _ _ __ __
l I INSERT TO ACTION a.1 OF SPECIFICATION 3.11.2.5, PAGE 3/4 11-15 Explanation of why gaseous radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems which resulted in gaseous radwaste being discharged without treatment, and the reason for the inoperability, , J 1 1 i I 4 J E68abm25
M OIDACTIVJ EFFLUENTS . ! 3/ u.3 50110 RADI0 ACTIVE WASTE I LIMIT C0kCITIONFbROPERATION - l i 3.11.3 The olisi radweste system as saecified in the PROCESS CONTROL ROGRAN l shall be GFE. *BLE and used, as applica)le in accordance'with a PROC CONTROL l ! PROGP.AM, for t.. SOLDIFICATION and packag ng of radioactive wastes o ensure ; i meetin.; the raq 'raruants of 10 CFR Part 2 and of 10 CFR Part 71 ior to l l shipacqt (.f redie etive wa:ites from the site. l APPLIQBILITY: 4t 1 times. . l ACT104-- l j l a. Wiin itia packag tg requirments of 10 CFR Part 0 and/or 10 CFR l 1 Part 71 not, satis ied, suspend shipments of factively packaged i i solid radioactive stas from the site. - l
- b. With the solid ra
- lwas system inoperabl for more than 31 consecutive' '
- drn , in lieu of any o er report requ' ed by 5 cification 6.9.1, p.v,;t.rt and subait to th Commission ithin 30 ys pursuant to
~
i j Soet:lfication 6.!I.2 a Spe al Repo which includes the following
- Inforzation:
\ INSERT * .. . Ny 1. Idustification of the ino able equipment or subsystems and j the .eason far inoperabil , . . $ I 2. Action (s)tu.(entores re th inoperable equipment to OPERABLE I 4 status, I 3. A description of alternative ed for 50LDIFICATION and j packaging of rad active wastes, a
- 4. Suciary desc i tion of action (s) taken to prevent a recurrence.
l
- c. The provisons of ,pecifications 3.0.3, 3.0.4 6.9.1.11 are ,not
! applicable. , i SURVEILLANCE RFQUIR NTS - . i 4. U.3.1 The h id rad.#a1te system shall be demonstrated OPERABLE t least ! once per 92 d by:
- a. C rating the rolid radweste system at least once in the prov us
<. days in ar.cerdance with the PROCESS CONTROL PROGRAM, or l b. Verification of the existence of a valid contract for 50LIDIFICA N to be performed by a contractor in accordance with a PROCESS CONT , , PROGRAM.
GRAND GULF-UNIT 1 3/4 11-18 A w ,J . ,,s W. l __ ..___n__ .-u_____. _. . . _ _ _ _ _ . _ . _ -
- .....~~m. . - , _ . . . _ _ _ _ . .n - .....s.......a..~...
l
- e
. . i l
i RADIDACTIVE EFFLUENTS i 3/4.11.3 SOLID HAD10 ACTIVE WASTE < e , LIMITING CONDITION FOR OPERATION 3.11.3 The solid radwaste system shall be used in accordance with a PROCESS CONTROL PROGRAM to process wet radioactive wastes to meet shipping and burial ground requirements. APPLICABILITY: At all times. 3 ACTION: ) a. With the provisions of the PROCESS CONTROL PROGRAM not satisfied, ' suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.
- b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS l .l 4.11.3 The PROCESS CONTROL PROGRAM shall be used to verify the 1 SOLIDIFICATION
- of at least one representative test specimen from at least j every tenth batch of each type of wet radioactive waste.
%g
- a. If any test specimen fails to verify SOLIDIFICATION *, the q)
SOLIDIFICATION
- of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION
- parameters can be determined in accordance with the i PROCESS CONTROL PROGRAM, and a subsequent test verifies l
SOLIDIFICATION *. SOLIDIFICATION
- of the batch may then be resumed using the alternative SOLIDIFICATION
- parameters determined by the PROCESS CONTROL PROGRAM.
".. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION *, the PROCESS CONTROL PROGRAM shall provide for the i
- collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least 3 i
consecutive initial test specimens demonstrate SOLIDIFICATION *. The PROCESS CONTROL PROGRAM shall be modified as required, as provided in Specification 6.13, to assure SOLIDIFICATION
- of subsequent batches of waste.
*Except dewatering.
GRAND GULF-UNIT 1 3/4 11-18 Amendment No. l 264abm1
r,- - -
. . 1
'MADI0 ACTIVE EFFLUENTS SURVE!L NCE REQUIREMENTS (Continued) 4.11.3.2 The P ESS CONTROL PROGRAM shall be used to verify t SOLIDIFICATION of at *. east one re esentative test specimen from at least ev y tenth batch 4
of aach type of wet dioactive waste, e.g., filter sludges spent resins, evaporator bottoms, an odium sulfate solutions, categor' ed as "non-specific" waste, except Equipment F ter Sludge and Floor Filter udge wastes which are requires at least one repres tative test specimen fr at least every twentieth haten. For batches categorize as " specific" wast , sampling shall be as out-lined in the PROCESS CONTROL PRO .
- a. If any test specimen of "no peci c" waste fails to ve'rify SOLIDIFI-CATION, the SOLIDIFICATION of batch under test shall be suspended
'. until such time as additional specimens can be obtained, altern- m ative SOLIDIFICATION paramet s can e determined in accordance with 8'
' the PROCESS CONTROL PROGR , and a su equnt test verifies SOLIDIFI- D CATION. SOLIDIFICATION the batch ma then be resumed using the alternative SOLIDIFICA ON parameters dete.-ined by the PROCESS CONTROL PROGRAM.
- b. If the initial t specimen from a batch of'"non to verify 50L FICATION, the PROCESS CONTROL PROGR ecific" waste fails shall provide for the col ction.and testing of representative test ecimens from -
, each cons utive batch of the same type of wet waste unt a: least i
3 conse tive initial test specimens demonstrate.50LIDICAI N. The PROCE CONTROL PROGRAM shall be modified as required, as pro ed in . Spe fication 6.13.2, to assure SOLIDIFICATION of subsequent bat es o waste. i i . I i 1 l I l l
/
L h l GRAND GULF-UNIT 1 3/4 11-19 A . g M = a # , #s. l
p eutre ea rtas pure + Ren A ca w erH M !*" RADIQ4CTIVE EFFLUENTS 3/ 11.4 TOTAL DOSE IMITING COSDITION F0 'PERATION i 3.11.4 The dose r dose commitments ove 12 consecutive months o any member { of the public, e to releases of radi ctivity and radiation from uranium less than or eq fuel cycle so ces shall be limited t : a. l to 25 mrem to the total bo or any organ (except he thyroid), b. less an or equal to 75 mrem to he thyroid. APPLICA LITY: At all times. ACTIQ
- i
{ /a. With the calcul in liquid or ed doses from the rel se of radioactive mate als seous effluents excee ng twice the limits of pecifica-tion 3.11.1. .a, 3.11.1.2.b, 3.11.2 .a. 3.11.2.2.b, 3.11. l
.a or , 3.11.2.3.b in lieu of any other r port reuqired by Speci cation 6.9.1, i
prepare d submit a Special Rep t to the Director, Nuc ear Reactor
. Regulat n, U.S. Nuclear Regul ory Commission, Washin on, D.C. '
! 20555, ithin 30 days, which fines the corrective ion to be j take to reduce subsequent r eases to prevent recu ence and exceeding 4 th imits of Specificatio 3.11.4. This Special eport shall
,~ i lude an analysis whic estimates the radiatio exposure (i.e.,
ia ose) to a member of th public from uranium f cycle sources 3 including all effluen pathways and direct ra ation for a
, 12 consecutive month eriod that includes,t release (s) coverd by this report. If t estimated dose (s) exc ds the limits of Speci ica- C l tion 3.11.4, and ' the release conditio resulting in violation f
! 40 CFR 190 has t already been correct , the Special Report r all { i include a req st for a variance in a ordance with the provi ons of 40 CFR 19 and including the spec
- ied information of I 0.00(b).
- j. Submittal o the report is conside' d a timely request, an a variance is grante until staff action on te request is complete The varianc only relates to the li ts of 40 CFR 190, and es not apply any way to the requir ents for dose limitat" n of 10 CFR Part , as addressed in oth sections of this tech cal
- spe fication, l
j b. e provisions of Specif cations 3.0.3 and 3.0. are not applicable. j SURVE NCE REQUIREMENTS i l f 4 1.4 Dose Calculations / Cumulative dose contr utionsfromliquidany f l aseous effluents shall e determined in accor nce wth Specifications . 11.1.2, j .11.2.2, and 4.11.2. , and in accordance wi the 00CM. l l GRAND GULF- T1 4_11-20
~ -
RADIOACTIVE EFFLUENTS 3/4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 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 mrem to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 arem . APPLICABILITY: At all times. ACTION:
- a. With the calculated doses from the release of radioactive. materials in liquid or gaseous effluents exceeding twice the limits of Specification 3.11.1.2.a. 3.11.1.2.b , 3.11.2.2.a. 3.11.2.2.b.
3.11.2.3.a. or 3.11.2.3.b, calculations should be made including direct radiation contributions from the reactor units and from outside storage tanks to determine whether the above limits of Specification 3.11.4 have been exceeded. If such is the case, prepare and submit to the commission within 30 days, pursuant to Specification 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 Part 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent fk pathways and direct radiation, for the calendar, year that includes g 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 resulting in violation of 40 CFR Part 190 has not already been corrected, the Special Report shall include a request for a variance in accordance 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 Specification 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 Specifications 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 reactor units and from radwaste storage tanks shall be determined in accordance with l the methodology and parameters in the ODCH. This requirement is applicable only under conditions set forth in Specification 3.11.4.a. I 3/4 11-20 GRAND GULF-UNIT 1 A ***M*4" N: - i L
* ~ ~ - ~~ ~~~ " - - -
- 3/4.12 RADIOLOG8 CAL ENV2RONMENTAL MONITOR 8HG 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12.1-1. -
APPLICABILITY: At all times. ACTION:
- a. With the radiological environmental monitoring program not being conducted as specified in Table 3.12.1-1, prepare and submit to the Commission, .in the Annual Radiological Environmental Operating .:
Report per Specification 6.9.1.7, 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 effluent in an' environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12.1-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days deem. Er, k t': :.d e' t'e t "ected es'e-de T --+e- a Special Report that. %,ISA identifies the cause(s) for exceeding the limit (s) and defines the r'*M'1 corrective actions to be taken to reduce radioactive affluents so that the potential annual dose to'- ' "'- '" is less than the gh[ggg calendar year limits of Specification 3.11.1.2, 3.11.2.2 and 3.11.2.3. h pursuant to Specification 6.9.1.13.f. When more than one of the
#W radionuclides in Table 2.1" " are detected in the sampling medium, l$"
fus& E this report shall be submitted if: N 3. It. l-2. l concentration (1) reporting level (1) , concentration (2) reporting level (2) + **'> 1.0 When radionuclides other than those in Table 3.12.1-2 are detected i ' and are the result of clant affluents, this report shall be submitted f the potent' al annual dose to': . ind';f d::1 is equal to or greater E than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 j and 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 Report. I c. Annual Radiological EmilkarD: 5 Environmental Opergting$2ED
'ONtD:Y:N: 'e from one or l
her fore of the sample locations required by Table 3.12.1-1, r:::-e : e :e sy sp--t