ML20209J159
| ML20209J159 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/09/1986 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20209J095 | List: |
| References | |
| NUDOCS 8609160173 | |
| Download: ML20209J159 (58) | |
Text
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I.
Technical Specification Change Request No.112 It is requested that the following revision be made to the TMI-1 Technical Specifications:
Replace pages 1, iii, iv, vi, 1-3, 1-4, 1-6, 1-7, 1-8, 3-13, 3-22, 3-96, 3-99, 3-101, 3-102, 3-103, 3-104, 3-106, 3-111, 3-115, 3-117, 3-127, 4-3, 4-4, 4-7a, 4-87, 4-90, 4-91, 4-92, 4-93, 4-9 4, 4-9 5, 4-9 6, 4-9 7, 4-98, 4-99, 4-100, 4-101, 4-102, 4-103, 4-104, 4-105, 4-106, 4-107, 4-108, 4-109, 4-110, 4-111.
Eliminate pages 1-6a, 3-105, 3-105a, 4-112 through 4-122.
Add new page 1-4a.
Please note the following pages are affected by pending Tech. Spec.
Change Requests: 3-102, 3-103, 4-93, 4-94, 4-102, 4-103 and 4-105.
II.
Reason for Change This change is requested to make numerous changes to the radiological effluent Technical Specifications for TMI-1.
The changes are administrative; the changes are made to improve clarity, to make the Technical Specifications more specific to THI-1 without changing the intent and to be more consistent with the current guidance provided by the NRC.
In addition, some administrative changes were made to definitions in Section 1 of the Technical Specifications.
III.
Safety Evaluation Justifying Change The changes proposed in TSCR 112 are administrative. Most of the changes enhance the clarity of the Technical Specifications. The remaining changes reflect the current NRC guidance for operating reactors, provided in NUREG-0472, Rev. 3, "The Standard Radiological Effluent Technical Specifications for Pressurized Water Reactors,"
September 1982.
For ease of review, the changes requested will be discussed on a page-by-page basis.
Page 1-3 Definition subjects have been capitalized.
FSAP, figure numbers have been updated.
TS 1.5.2 Channel Test. The definition is deleted and replaced with a revised definition from TS 1.11.
The test for bistable channels has been eliminated because none exist at TMI-1.
8609160173 860909 PDR ADOCK 05000289
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Page 1-4 Definition subjects have been capitalized.
TS 1.5.3 Instrument Channel Check has been deleted and replaced with the STS definition of TS 1.10 Channel Check.
Page 1-4a has been added.
Page 1-6 TS 1.9-1.11 has been deleted and are superceded by TS 1.5.2-1.5.4.
TS 1.9 is unnecessary due to TS 1.5.4.
TS 1.12 has been revised to add an additional source for thyroid dose conversion factors, Table E-7 of Reg. Guide 1.109. This is consistent with STS.
TS 1.15 has been revised to clarify that the 0D04 is also used in the conduct of the environmental radiological monitoring program.
TS 1.17 has grammatical changes.
Page 1-7 TS 1.20 has been changed to delete the unnecessary phrase "as required during VENTING."
Page 1-8 The notation S/U has been changed to PS/U and the frequency expanded to include 7 days prior to reactor startup. This is a more restrictive and better defined frequency interval.
The frequency for a refueling interval of or.ce per 18 months is being deleted.
A refueling interval is defined in TS 1.2.8 as not to exceed 24 months, therefore, for consistency, the frequency is removed.
A new notation "E" with the frequency of once per 18 months is being added. Table 1.2, when initially proposed, was to apply to those tables which are part of the Radiological Effluent Technical Specifications. A frequency specified was once per refueling interval, defined as once per 18 months.
This created an inconsistency in the definition of refueling interval as noted above.
The notation "E" will now be used when the frequency is to be once per 18 months.
Page 3-13 Defined terms have been capitalized.
Footnote (a) has been changed to correct a typographical error in the Specification identified.
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t Page 3-22 Defined terms have been capitalized.
Page 3-96 Defined terms have been capitalized and the convention used for valves has been made consistent with what is currently used.
Page 3-99 For clarity, Table 3.21-1 Action 18 (1.) has been revised to ensure samples are analyzed in accordance with TS 4.22.1.1 A and B.
Action 18 (3.) has been revised to correct the title given.
The 14-day release limit has been eliminated from Action 18. This is consistent with HUREG-0472, Rev. 3, September 1982.
Upon exceeding the 14-day limit, the action would have been to suspend the release of radioactive effluent. This is unnecessary considering the relative importance of the instrumentation and the adequacy of grab sampling.
The elimination of the time limit constitutes a clarification.
Action 20 has been revised to eliminate the 30-day release limit and extend the sampling frequency to once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during release. This is consistent with NUREG-0472, Rev. 3, September 1982.
Upon exceeding the 30-day release limit, the action would be to place the plant in hot shutdown. The rationale for eliminating the time limit is the same as Action 18 above. The extension of time between samples still constitutes an acceptable sampling frequency. Also, the requirement to analyze for both gross beta and gamma has been revised to require beta or gamma analysis. This is consistent with NUREG-0472, Rev. 3, September 1982.
Action 21 has been revised to eliminate the 30-day release limit. This is consistent with NUREG-0472, Rev. 3, September 1982. The rationale for eliminating the time limit is the same as Action 18.
Page 3-101 The items from page 3-102 have been moved to this page.
The
- symbol for applicability has been changed to # for item 3 and defined as "At all times during containment purging" in the Table Notation.
This is for clarification. The instrumentation is used only during containment purging, therefore, the action statements are only meaningful during containment purging. 1
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Page 3-102 Items from page 3-103 and 3-104 have been moved to this page.
The symbol
- for applicability has been changed to ## for item 4 and defined as "At all times when condenser vacuum is established" in the Table Notation. This is for clarification.
The condenser vent system instrumentation should only be required to be operable when the system is in use.
The measuring devices for 5.d have been specified for i
N clarity.
Page 3-103 Actions 25-27 have been moved from page 3-105 to this page.
Notations # and ## have been added as explained above.
Action 25 has been revised to eliminate the 14-day 1
release limit.
This is consistent with NUREG-0472, Rev. 3, September 1982. Upon exceeding the 14-day limit, the action would be to suspend release of radioactive effluent. This is unnecessary considering the relative importance of the instrumentation and the adequacy of grab sampling.
The elimination of the time limit constitutes a clarification.
t For clarity, Action 25 (1.) has been revised to ensure samples are analyzed in accordance with Table 4.22-2, Item A.
Action 26 has been revised to eliminate the 28-day release limit.
This is censistent with NUREG-0472, Rev. 3, September 1982.
Upon exceeding the 28-day release limit, the action would he to place the plant in hot shutdcwn. The rationale for eliminating the time ifmit is the same as for Action 25.
Action 27 has been revised to eliminate the 28-day release ifmit and extend the sampling frequency to once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This is consistent with NUREG-0472, Rev. 3, September 1982.
The rationale for eliminating the time is the same as for Action 25. The extension of time between samples still constitutes an acceptable sampling frequency.
The phrase "af ter the channel has been declared inoperable" has been added to Action 27 for clarity.
The note in parenthesis in Action 27 has been clarified to show it applies to RM-A9..
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Page 3-104 Action 30-32 have been moved to this page.
Action 31 has been revised to eliminate the 28-day release limit. This is consistent with NUREG-0472, Rev. 3, September 1982.
The rationale for eliminating the time limit is the same as for Action 25 except that auxiliary sampling equipment is relied on instead of grab sampling.
Page 3-106 Action a has been revised by changing " site" to " unit" for clarity.
Page 3-111 TS 3.22.2.1.b has been revised to specify what is included in the Specification.
This is consistent with NUREG-0472, Rev. 3, September 1982.
Page 3-115 The reference to TS 6.9.2 reporting requirements has been deleted. The reporting requirement is now consistent with the rule changes to 10 CFR 50.72 and 50.73.
The 31-day time periods have been changed to monthly time periods.
This is part of the attempt to achieve consistency in format throughout the Tech. Specs.
Page 3-117 TS 3.22.2.6 has been retitled to Waste Gas Decay Tanks.
This is the proper name for what were called Gas Storage Ta nks.
Identifier "a." has been added to the Action Statement.
Page 3-127 Identifier "a." has been added to the Action Statement.
Page 4-3 The test frequency notation for items 5 and 6 have been changed to PS/U to be consistent with lable 1.2.
Page 4-4 Items 8 and 9 have been moved from page 4-3 to this page.
Page 4-7a The notation key has been deleted. Table Notations are all listed in Table 1.2.
Page 4-87 TS 4.21 has been retitled for clarity.
Page 4-90 TS 4.21.2 has been retitled for clarity.
The title of Table 4.21-2 has been corrected. -
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Page 4-9 Channel Calibration frequency notation "R" has been changed to "E" for consistency with Table 1.2.
Items 3 and 4 have been moved to this page.
The applicability notation of items 3.a. through e. has been changed from
- to # and defined as "at all times during containment purging" in the Table Notation.
This is for clarification. The instrumentation is used only during containment purging, therefore, the surveillance requirements are only meaningful during containment purging.
The applicability notation of item 4.a. has been changed from
- to ## and redefined in the Table Notation as "at all times when condenser vacuum is established." This is for clarification. The instrument is used only when the system is in use, therefore, the surveillance requirements are only meaningful when the system is in use.
The measuring device in Item 3.d has been specified.
Item 4.a has been revised to include the suitable equivalent to RM-A5 in the surveillance to ensure operability of the equivalent monitor.
The title of Table 4.21-2 has been corrected.
Page 4-92 Item 5 has been moved to this page.
Channel Calibration frequency notation "R" has been changed to "E" for consistency with Table 1.2.
The quarterly channel test for item 5.e., the sample flow rate monitor for the Auxiliary and Fuel Handling Building Ventilation System, has been eliminated.
This test is of no value because this type of flow rate monitor does not provide input to any channel.
The measuring devices in Item 5.d have been specified.
Page 4-93 The notations # and ## have been added.
The Table Notation for Table 4.21-2 has been moved to this page.
Page 4-94 Itent (5) has been moved to this page.
Page 4-95 TS 4.22.1 has oeen titled " Liquid Effluents" and moved from page 4-97.
TS 4.22.1.1 has been titled " Concentration."
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t Page 4-95 The designatnrs A, B and C have been added.
(Conti nued)
A grammatical change was made in TS 4.22.1.1.C.
Page 4-96 This information has been moved from page 4-98.
Table 4.22-1, Item A.1 has been revised to delete gross beta and P-32 activity analysis and Table Notation (g).
Notation (g) is incorrectly referenced since it does not involve gamma emitters.
The deletion of gross beta and P-32 is consistent with NUREG-0472, Rev. 3, September, 1982.
Gross beta consideration is not required and P-32 has been found not to be a major contributor to dose from liquid effluents.
Page 4-97 This information has been moved from page 4-99.
Item A.2 has been revised to delete P-32 activity analysis and Table Notation (g) as justified above.
Page 4-98 Table Notations (a) and (b) have been moved from page 4-100 to this page.
Table Notation (a) has been revised to correct 106 to 6
10,
Page 4-99 Table Notations (c) through (f) have been moved to this page.
Table Notation (g) has been deleted because it is no longer referenced in Table 4.22-1.
Page 4-100 The specifications from pages 4-102 through 4-104 have been moved to this page.
TS 4.22.1.3 has been titled " Liquid Waste Treatment."
The time period of "once per 31 days" has been replaced with "once a month". This is to achieve consistency in format in the Tech. Specs.
TS 4.22.1.4 has been titled " Liquid Holdup Tanks." The word "above" has been deleted.
Page 4-101 The specifications from page 4-105 have been moved to this page.
TS 4.22.2 has been titled " Gaseous Effluents."
TS 4.22.2.1 has been titled " Dose Rates."
TS 4.22.2.1.2 now references TS 3.22.2.lb. _
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Page 4-102 Table 4.22-2 now begins on page 4-102 and the title nas been corrected to " gaseous waste."
The requirement to analyze for I-133 in Item D has been eliminated.
This is not required per NUREG-0472, Rev. 3, September 1982.
H-3 activity analysis and the associated lower limit of detection have been added for containment purge release analysis and the redundant analysis deleted from the Auxiliary and Fuel Handling Building Ventilation Analysis.
This corrects an error made when the table was originally prepared.
The analysis of H-3 is being performed prior to containment purge.
Item A has been retitled " Waste Gas Decay Tank" for consistency with current terminology.
Page 4-103 The title of Table 4.22-2 has been corrected.
Page 4-104 Table Notation (b) has been revised to be consistent w'ith NUREG-0472, Rev. 3, September 1982.
Page 4-105 Notation (d) has been revised to identify the samples to be changed as charcoal cartridges and particulate fil ters. This is for clarification.
Notation (e) has been revised to delete "from the ventilation exhaust." A ventilation sample point does not exist in the spent fuel pool area, however, grab samples are taken, which is an equally conservative sampling method.
Notation (h) has been revised to be consistent with NUREG-0472, Rev. 3, September 1982.
Page 4-106 The surveillance requirements from pages 4-110 through 4-115 has been moved to this page.
TS 4.22.2.2 through TS 4.22.2.6 have been retitled.
TS 4.22.2.4.2 has been eliminated.
This is consistent with NUREG-0472, Rev. 3, September,1982.
In TS 4.22.2.4, the " site" has been changed to the " unit" for clarity.
The 31-day time period has been changed to monthly as an attempt to achieve consistency in format throughout the Tech. Specs. _
t Page 4-106 TS 4.22.2.5 has been revised to define the limits as (Continued) those in TS 3.22.2.5.
Also, " required OPERABLE by" has been changed to " covered in" and the requirement to continucusly monitor has been deleted. This is to provido consistency with the applicability specified in Table 3.21-2, Item 2.
TS 4.22.2.6 has been revised to specify the tank as the waste gas decay tank and the limit as < 8800 ci/ tank.
Page 4-107 These items were moved from page 4-115 to 4-107.
TS 4.22.3 has been titled " Solid Radioactive Waste."
TS 4.22.3.1.1 and TS 4.22.3.1.2 have been titled and renumbered as TS 4.22.3.1 and TS 4.22.3.2.
Page 4-108 TS 4.22.4 has been titled and moved from page 4-116 to 4-108.
Page 4-109 Pages 4-117, 4-121 and 4-122 have been consolidated on page 4-109.
TS 4.23 has been titled " Radiological Environmental Mo nito ri ng. " TS 4.23.1 through 4.23.3 has been titled.
Grammatical changes were made in TS 4.23.1.
TS 4.23.2 has been revised to read the census shall be conducted at least once per 12 months.
The methods of gathering information have been made less specific.
These changes do not change the intent of the Specification, but allow a more effective program to be implemented.
TS 4.23.3 has been revised to eliminate the meaningless phrase "above required." The phrase "in accordance with the ODCM" has been deleted. The Inter-Laboratory Comparison Program is not included in the ODCM. This is an administrative change.
Page 4-110 Table 4.23-1 has been moved from page 4-118 to 4-110.
The notation for the LLD for water in 131 I analysis has been changed to (b). This is correcting a typographical erro r.
137 s analysis har been changed to The LLD for milk for C
- 18. This is correcting a typographical error.
Page 4-111 Pages 4-119 and 4-120 have been consolidated on page 4-111.
Changes have been made to notation (a). These changes are consistent with NUPEG-0472, Pev. 3, September 1982.
The existing notation does not apply to environmental samples.
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IV.
No Significant Hazards Consideration The proposed changes provide clarity and consistency with the current NRC guidance and:
1.
does not affect plant design or operation, and therefore would not involve a significant increase in the probability or consequences of an accident previously evaluated; 2.
does not involve modification to existing plant equipment, and therefore would not create the possibility of a new or different kind of accident from any accident previously evaluated; 3.
does not involve changes which would affect the safety analysis of the plant, and therefore would not involve a significant reduction in a margin of safety.
The proposed amendment combines Example (1) and Example (vii) of amendments that are consicared not likely to involve significant hazards consideration (48 FR 14870) in that the changes are purely administrative or are changes to conform with the current NRC guidance which result in minor changes to operations clearly in keeping with the Standard Radiological Effluent Technical Specifications, NUREG-0472, Rev. 3.
V.
Implementation It is requested that this amendment become effective upon issuance and 60 days be allowed for implementation of any procedure changes that are required.
VI.
Amendment Fee (10 CFR 170.21)
In accordance with the provisions of 10 CFR 170.21, a check for $150.00 is enclosed.
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TABLE OF CONTENTS Section Page TECHNICAL SPECIFICATIONS 1
DEFINITIONS 1 -1 l.1 RATED POWER l -1 1.2 REACTOR OPERATING CONDITIONS 1 -1 1.2.1 Cold Shutdown 1 -1 1.2.2 Hot Shutdown 1 -1 1.2.3 Reactor Critical 1 -1 1.2.4 Hot Standby 1 -1 1.2.5 Power Operation 1 -1 1.2.6 Refueling Shutdown 1 -1 1.2.7 Refueling Operation 1-2 1.2.8 Refueling Interval 1-2 1.2.9 Startup l-2 T vg l-2 1.2.10 A
1.2.11 Heatup-Cooldown Mode 1-2' l.2.12 Station, Unit, Plant, and Facility 1-2 1.3 OPERABLE l-2 1.4 PRUTECTIVE INSTRUMENTATION LOGIC 1-2 1.4.1 Instrument channel 1-2 1.4.2 Reactor Protection System 1-2 1.4.3 Protection Channel 1-3 1.4.4 Reactor Protection System Logic l-3 1.4.5 Engineered Safety Features System 1-3 1.4.6 Degree of Redundancy 1-3 1.5 INSTRUMENTATION SURVEILLANCE l-3 1.5.1 Tii) Test 1-3 1.5.2 Channel Test 1-3 1.5.3 Channel Check 1-4 1.5.4 Channel Calibration 1-4 1.5.5 Heat Balance Check 1-4 1.5.6 Heat Balance Calibration 1-4 1.6 POWER DISTRIBUTION 1-5 1.6.1 Quadrant P6wer Tilt 1-5 1.6.2 Reactor Power Imbalance 1-5 1.7 CONTAINMENT INTEGRITY l-5 1.8 FIRE SUPRESSION WATER SYSTEM l-5
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1.12 DOSE EQUIVALENT I-131 1-6 1.13 SOURCE CHECK 1-6 1.14 SOLIDIFICATION 1-6 1.15 0FFSITE DOSE CALCULATION MANUAL l-6 1.16 PROCESS CONTROL PROGRAM l-6 1.17 GASEOUS RADWASTE TREATMENT SYSTEM 1-6 1.18 VENTILATION EXHAUST TREATMENT SYSTEM l-6 1.19 PURGE-PURGING l-7 1.20 VENTING l-7 i
Amendment No.11, 72
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TABLE OF CONTENTS Section Page 3.16 SHOCK SUPPRESSORS (SNUBBERS) 3-63 3.17 REACTOR BUILDING AIR TEMPERATURE 3-80 3.18 FIRE PROTECTION 3-86 3.18.1 Fire Detection Instrumentation 3-86 3.18.2 Fire Suppression Water System 3-88 3.18.3 Deluge / Sprinkler Systems 3-89 3.18.4 C02 System 3-90 3.18.5 Halon Systems 3-91 3.18.6 Fire Hose Stations 3-92 3.18.7 Fire Barrier Penetration Seals 3-94 3.19 CONTAINMENT SYSTEMS 3-95 3.20 SPECIAL TEST EXCEPTIONS 3-95a 3.20.1 Low Power Natural Circulation Test 3-95a 3.21 RADI0 ACTIVE EFFLUENT INSTRUMENTATION 3-96 3.21.1 Radioactive Liquid Effluent Instrumentation 3-96 3.21.2 Radioactive Gaseous Process and Effluent Monitoring 3-100 Instrumentation 3.22 RADI0 ACTIVE EFFLUENTS 3-106 l
3.22.1 Liquid Effluents 3-106 3.22.2 Gaseous Effluents 3-111 3.22.3 Solid Radioactive Waste 3-118 3.22.4 Total Dose 3-119 3.23 RADIOLOGICAL ENVIRONMENTAL MONITORING 3-120 l
3.23.1 Monitoring Program 3-120 3.23.2 Land Use Census 3-125 3.23.3 Interlaboratory Comparison Program 3-127 4
SURVEILLANCE STANDARDS 4-1 4.1 OPERATIONAL SAFETY REVIEW 4-1 4.2 REACTOR COOLANT SYSTEM INSERVICE INSPECTION 4-11 4.3 TESTING FOLLOWING OPENING OF SYSTEM 4-28 4.4 REACTOR BUILDING 4 -29 4.4.1 Containment Leakage Tests 4-29 4.4.2 Structural Integrity 4-35 4.4.3 Deleted 4-37 4.5 EMERGENCY LOADING SEQUENCE AND POWER TRANSFER, EMERGENCY 4 -39 CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 Emergency Loading Sequence 4-39 4.5.2 Emergency Core Cooling System 4-41 4.5.3 Reactor Building Cooling and Isolation System 4-43 4.5.4 Decay Heat Removal System Leakage 4-45 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTS 4-46 i
111 Amendment No. 72, 81, 108
J TABLE OF CONTENTS i
Section Pac _ e 4.7 REACTOR CONTROL R0D SYSTEM TESTS 4-48 4.7.1 Control Rod Drive System Functional Tests 4-48 4.7.2 Control Rod Program Verification 4-50 4.8 MAIN STEAM ISCLATION VALVES 4-51 4.9 EMERGENCY FEEDWATER PUMPS PERIODIC TESTING 4-52 4.9.1 Test 4-52 4.9.2 Acceptance Criteria 4-52 4.10 REACTIVITY ANOMALIES 4-53 4.11 REACTOR COOLANT SYSTEM VENTS 4-54 4.12 AIR TREATMENT SYSTEMS 4-55 4.12.1 Emergency Control Room Air Treatrent System 4-55 4.12.2 Reactor Building Purge Air Treatment System 4-55b 4.12.3 Auxiliary and Fuel Handling Exhaust Air Treattent System 4-55d 4.13 RADI0 ACTIVE MATERIALS SOURCES SURVEILLANCE 4-56 4.14 DELETED 4-57 4.15 MAIN STEAM SYSTEM INSERVICE INSPECTION 4-58 4.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE 4-59 4.17 SHOCK SUPPRESSORS (SNUBBERS) 4-60 4.18 FIRE PROTECTION SYSTEMS 4-72 4.18.1 Fire Protection Instruments 4-72 4.18.2 Fire Suppression Water System 4-73 4.18.3 Deluge / Sprinkler System 4-74 4.18.4 CO2 System 4-74 4.18.5 Halon Systems 4-75 4.18.6 Hose Stations 4-76 4.18.7 Fire Barrier Penetration Seals 4-76a l
4.19 OTSG TUBE INSERVICE INSPECTION 4-77 4.19.1 Steam Generator Sample Selection and Inspection Methods 4-77 4.19.2 Steam Generator Tube Sample Selection and Inspection 4-77 4.19.3 Inspection Frequencies 4-79 4.19.4 Acceptance Criteria 4-80 4.19.5 Reports 4-81 4.20 REACTOR BUILDING AIR TEMPERATURE 4-86 4.21 1%UTUT3TIVE EFFLUENT INSTRUMENTATION 4-87 l
4.21.1 Radioactive Liquid Effluent Instrumentation 4-87 4.21.2 Radioactive Gaseous Process and Effluent Monitoring 4-90 Instrumentation 4
i 4.22 RADI0 ACTIVE EFFLUENTS 4-95 4.22.1 Liquid Effluents 4-95 4.22.2 Gaseous Effluents 4-101 4.22.3 Solid Radioactive Waste 4-107 4.22.4 Total Dose 4-108 4.23 RADIOLOGICAL ENVIRONMENTAL MONITORING 4 -1 09 4.23.1 Monitoring Program 4-1 09 4.23.2 Land Use Census 4 -1 09 4.23.3 Interlaboratory Comparison Program 4 -1 09 iv Amendment No. 11, 28, 30, 41, 47, 55, 72, 78, 95, 97
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LIST OF TABLES TABLE TITLE PAGE 1.2 Frequency Notation 1-8 2.3-1 Reactor Protection System Trip Setting Limits 2-9 3.1.6.1 Pressure Isolation Check Valves Between the Primary 3-15a Coolant System and LPIS
- 3. 5-1 Instruments Operating Conditions 3-29 3.5-2 Accident Monitoring Instrements 3-40c 3.5-3 Post Accident Monitoring Instrumentation 3-40d 3.18-1 Fire Detection Instruments 3-87 3.21 -1 Radioactive Liquid Effluent Monitoring Instrumentation 3-97 3.21 -2 Radioactive Gaseous Process and Effluent 3-101 Monitoring Instrumentation 3.23-1 Radiological Environmental Monitoring Program 3-122 3.23-2 Reporting Levels for Radioactivity Concentration 3-126 in Environmental Samples 4.1 -1 Instrument Surveillances Requirements 4-3 4.1 -2 Minimum Equipment Test Frequency 4-8 4.1 -3 Minimum Sampling Frequency 4-9 4.1 -4 Post Accident Monitoring Instrumentation 4-10a 4.19-1 Minimum Number of Steam Generators to be 4-84 Inspected During Inservice Inspection 4.19-2 Steam Generator Tube Inspection 4-85 4.21 -1 Radioactive Liquid Effluent Monitoring 4-88 Instrumentation Surveillance Requirements 4.21 -2 Radioactive Gaseous Effluent Monitoring 4-91 Instrumentation Surveillance Requirements 4.22-1 Radioactive Liquid Waste Sanpling & Analysis Proc * -
4-96 4.22-2 Radioactive Gaseous Waste Sampling & Analysis Program 4-102 4.23-1 Maximum Values for the Lower Limits of Detection (LLD) 4-110 vi Amendment No. 59, 72, 100, 106
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a forms the automatic system that protects the reactor by control rod trip.
It includes the four protection channels, their associated instrument channel inputs, manual trip switch, all rod drive control protection trip breakers, and activating relays or coils.
1.4.3 PROTECTION CHANNEL A PROTECTION CHANNEL as shown in Figure 7.1-1 of the updated FSAR (one of three or l
one of four independent channels, complete with sensors, sensor power supply units, amplifiers, and bistable modules provided for every reactor protection safety parameter) is a combination of instrument channels forming a single digital output to the protection system's coincidence logic.
It includes a shutdown bypass circuit, a protection channel bypass circuit and a reactor trip module.
1.4.4 REACTOR PROTECTION SYSTEM LOGIC This system utilizes reactor trip module relays (coils and contacts) in all four of the protection channels as shown in Figure 7.1-1 of the updated FSAR, to provide l
reactor trip signals for de-energizing the six control rod drive trip breakers.
The control rod drive trip breakers are arranged to provide a one-out-of-two-times-two logic.
Each element of the one-out-of-two-times-two logic is controlled by a separate set of two-out-of-four logic contacts from the four reactor protection channels.
1.4.5 ENGINEERED SAFETY FEATURES SYSTEM This system utilizes relay contact output from individual channels arranged in three analog sub-systems and two two-out-of-three logic sub-systems as shown in Figure 7.1-4 of the updated FSAR. The logic sub-system is wired to provide l
appropriate signals for the actuation of redundant engineered safety features equipment on a two-of-three basis for any given parameter.
1.4.6 DEGREE OF REDUNDANCY The difference between the number of operable channels and the number of channels which, when tripped, will cause an automatic system trip.
1.5 INSTRUMENTATION SURVEILLANCE 1.5.1 TRIP TEST A TRIP TEST is a test of logic elements in a protection channel to verify their associated trip action.
1.5.2 CHANNEL TEST A CHANNEL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practical to verify OPERABILITY, including alarm and/or trip functions.
1-3 L -
o 1.5.3 CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrumentation channels measuring the same parameter.
1.5.4 CHANNEL CALIBRATION An instrument CHANNEL CALIBRATION is a test, and adjustment (if necessary), to establish that the channel output responds with acceptable range and accuracy to known values of the parameter which the channel measures or an accurate simulation of these values. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include the channel test.
1.5.5 HEAT BALANCE CHECK A HEAT BALANCE CHECK is a comparison of the indicated neutron power and core I
thermal power.
I 1.5.6 HEAT BALANCE CALIBRATION A HEAT BALANCE CALIBRATION is an adjustment of the power range channel amplifiers I
output to agree with the core thermal power as defined by a weighted primary and 3
secondary heat balance considering heat losses. The weighting factor, a is shown in the figure below as a function of power level.
The equations below define the value of a as a function of power level and the use of a in determining the core thermal power.
1.0 o
58 EUs m "<
0 I
0 POWER 1-4 1
Core Thermal Power = a Qsec + (1-a ) Qprim for, POWER less than or equal to 15%, a=0 POWER greater than 15% AND less than 50%
a POWER - 15
=
85 J
WHERE:
POWER =
(Iprim 100 j
(max
' POWER greater than 50% AND less than 100%
POWER - 15 WHERE: POWER = Qsec 100 i
a
=
~
85 Qmax POWER greater than or equal to 100%,
a=1 1
1 1-4a 4
->.n.,r--
n-
, -, - ~ -,.,,
--.,e
-m
-,,-r
.---,,,--,,,,.w
,m
~
e-
,v, e--,,-
x,- -,,,,,-, -, -, -
E a
(Definitions 1.9 - 1.11 have been deleted).
1.12 DOSE EQUIVALENT I-131 The DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID 14844, " Calculation of Distance Factors for Power and Test Reactor Sites". [0r in Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977.]
1.13 SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
1.14 SOLIDIFICATION SOLIDIFICATION shall be the conversion of radioactive wastes from liquid treatment systems to a uniformly distributed, monolithic immobilized solid with definite volume and shape, bocaied by a stable surface of distinct outline on all sides (f ree-standi ng).
1.15 0FFSITE DOSE CALCULATION MANUAL (ODCM)
An 0FFSITE DOSE CALCULATION MANUAL (0DCM) shall be a manual containing the methodology and parameters to be used in the calculation of off-site doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints, and in the conduct of the environmental radiological monitoring program.
1.16 PROCESS CONTROL PROGRAM (PCP)
The PROCESS CONTROL PROGRAM shall contain the sampling, analysis, and formulation determination by which SOLIDIFICATION of radioactive wastes from liquid systems is assured.
1.17 GASEOUS RADWASTE TREATMENT SYSTEM The GASE0US RADWASTE TREATMENT SYSTEM is the system designed and installed to l
reduce radioactive gaseous effluents by collecting primary coolant system off gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
1.18 VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the purpose of removing fodines or particulates from the gaseous exhaust system prior to the release to the environment.
Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEMS.
1-6 Amendment No. 72
o 1.19 PURGE - PURGING PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions in such a manner that replacement air or gas is required to purify the confinement.
1.20 VENTING VENTING is the controlled process of discharging air as gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating conditions in such a manner that replacement air or gas is not provided. Vent used in system name does not imply a VENTING process.
I l
l-7 Amendment No. 72
-. +.
,,._.,,_-..,,-..,,.,__,n_,_,.,__.___,.
TABLE 1.2 FRE0VENCY NOTATION NOTATION FREQUENCY S
Shif tly (once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />)
D Daily (once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)
W Weekly (once per 7 days)
M Monthly (once per 31 days)
Q Quarterly (once per 92 days)
S/A Semi-Annually (once per 184 days)
R Refueling Interval P S/U Prior to each reactor startup, if not done during the previous 7 days P
Completed prior to each release N/A (NA)
Not applicable E
Once per 16 months The Surveillance Requirements shall be performed within the specified time interval with:
A.
A maximum allowable extension not to exceed 25% of the surveillance interval, and B.
A total maximum combined interval time for any 3 consecutive l
tests not to exceed 3.25 times the specified surveillance interval.
i i
i l
1-8 l
Amendment No. 72 l
3.1. 6.9 Loss of reactor coolant through reactor coolant pump seals and system valves to connecting systems which vent to the gas vent header and from which coolant can be returned to the reactor coolant system shall not be considered as reactor coolant leakage and shall not be subject to the consideration of Specifications 3.1. 6.1, 3.1. 6. 2, 3.1. 6. 3, 3.1. 6. 4, 3.1.6.5, 3.1.6.6 or 3.1.6.7, except that such losses when added to leakage shall not exceed 30 gpm.
If leakage plus losses exceeds 30 gpm, the reactor shall be placed in HOT SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection. l 3.1.6.10 Operating conditions of POWER OPERATION, STARTUP and HOT SHUTDOWN apply to the operational status of the high pressure isolation valves between the primary coolant system and the low pressure injection system.
a.
During all operating conditions in this specification, all pressure isolation valves listed in Table 3.1.6.1 that are located between the primary coolant system and the LPIS shall function as pressure isolation devices except as specified in 3.1.6.10.b.
Valve leakage shall not exceed the amount indicated in Table 3.1.6.1.(a) b.
In the event that integrity of any high pressure isolation check valves specified in Table 3.1.6.1 cannot be demonstrated, reactor operation may continue provided that at least two valves in each high pressure line having a non-functional valve are in and remain in, the mode corresponding to the isolated condition. (b) c.
If Specification 3.1.6.10.a or 3.1.6.10.b cannot be met, an orderly shutdown shall be accomplished by achieving HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Bases Any leak of radioactive fluid, whether from the reactor coolant system primary boundary or not, can be a serious problem with respect to in-plant radioactive contamination and required cleanup or, in the case of reactor coolant, it could develop into a still more serious problem and, therefore, the first indications of such leakage will be followed up as soon as practical. The unit's makeup system has the capability to makeup considerably more than 30 gpm of reactor coolant leakage.
Water inventory balances, monitoring equipment, radioective tracing, boric acid crystalline deposits, and physical inspections can disclose reactor coolant leaks.
l l
(a)
For the purpose of this specification integrity is considered to have been demonstrated by meeting Specificetion 4.2.7.
l l
(b)
Motor operated valves shall be placed in the closed position and power supplies deenergized.
i 3-13 Order dtd. 4/20/81 i
e e
e.
Core flood tank (CFT) vent valves CF-V3A and CF-V3B shall be closed and the breakers to the CFT vent valve motor operators shall be tagged open, except when adjusting core flood tank level and/or pressure. Specification 3.0.1 opplies.
3.3.1.3 Reactor Building Spray System and Reactor Building Emergency Cooling System The following components must be OPERABLE:
l a.
Two reactor building spray pumps and their associated spray nozzles headers and two reactor building emerigency cooling fans and associated cooling units (one in each train).
Specification 3.0.1 applies.
b.
The sodium hydroxide (Na0H) tank shall be maintained at 8 ft.
6 inches lower than the BWST level as measured by the BWST/Na0H tank differential pressure indicator. The NaOH tank concentration shall be 10.0
.5 weight percent (%).
c.
All manual valves in the discharge lines of the sodium hydroxide tank shall be locked open.
3.3.1.4 Cooling Water Systems - Specification 3.0.1 applies.
a.
Two nuclear service closed cycle cooling water pumps must be OPERABLE.
l b.
Two nuclear service river water pumps must be OPERABLE.
l c.
Two decay heat closed cycle cooling water pumps must be OPERABLE.
l d.
Two decay heat river water pumps must be OPERABLE.
e.
Two reactor building emergency cooling river water pumps must be OPERABLE.
3.3.1.5 Engineered Safeguards Valves and Interlocks Associated with the Systems in Specifications 3.3.1.1, 3.3.1.2, 3.3.1.3, 3.3.1.4 are OPERABLE. Specification 3.0.1 applies.
3.3.2 Maintenance shall be allowed during power operation on any component (s) in the makeup and purification, decay heat, RB emergency cooling water, RB spray, CFT pressure instrumentation, CFT level instrumentation, BWST level instrumentation, or cooling water systems which will not remove more than one train of each system from service. Components shall not be removed from service so that the affected system train is inoperable for more than 72 consecutive hcurs.
If the system is not restored to meet the requirerrents of Specification 3.3.1 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the reactor shall be placed in a COLD SHUTDOWN condition within twelve hours.
3-22 Amendment No. 33, 80, 98
3.21 RADIOACTIVE EFFLUENT INSTRUMENTATION 3.21.1 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.21.1 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.21-1 shall be OPERABLE with their alarm / trip setpoints set to ensure l
that the limits of Specification 3.22.1.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined in accordance with the Offsite Dose Calculation Manual (0DCM).
APPLICABILITY:
At all times
- ACTION:
a.
With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel i noperable, b.
With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.21 -1.
- For FT-84, and RM-L6, operability is not required when discharges are positively controlled through the closure of WDL-V257.
- For RM-L12 and associated IWTS/IWFS flow interlocks, operability is not required when discharges are positively controlled through the closure of IW-V72, 75 and IW-V280, 281.
- For FT-146, operability is not required when discharges are positively controlled through the closure of WDL-V257, IW-V72, 75 and IW-V280, 281 BASES t
The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases. The alarm / trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm / trip will. occur prior to exceeding the limits of 10 CFR Part 20.
3-96 Amendment No. 72, 88
--.m
TABLE 3.21-1 (continued)
TABLE NOTATION ACTION 18 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases may be resumed, provided that prior to initiating a. release:
1.
At least two independent samples are analyzed in accordance with Specifications 4.22.1.lA & B and; 2.
At least two technically qualified members of the Unit staff independently verify the release rate calculatfons and verify the discharge valve lineup.
3.
Operations and Maintenance Director Unit 1 shall approve each release.
Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 20 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that grab samples are collected and analyzed for gross radioactivity (beta or gamma) at a limit of detection if at least l
lx10-7 microcuries/ml, prior to initiating a release and at least ance per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during release.
ACTI0rl 21 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, radioactive effluent releases via this pathway may continue, provided the flow rate is estimated at least once l
per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases.
Pump curves may be used to estimate flow.
i t
3-99 Amendment No. 72, 88
E k
TABLE 3.21-2 9
k RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION E
h MINIMUM CHANNELS INSTRIMENT OPERABLE APPLICABILITY ACTION
-o" 1.
Waste Gas Holdup System a.
Noble Gas Activity Monitor 1
25 (RM-A7) b.
Effluent System Flow Rate 1
26 Measuring Device (FT-123) 2.
Waste Gas Holdup System Explosive Gas Monitoring System a.
Hydrogen Monitor 2
30
~
b.
Oxygen Monitor 2
30 S
3.
Containment Purge Monitoring System a.
Noble Gas Activity Monitor 1
27 (RM-A9 )
b.
Iodine Sampler (RM-A9) 1 31 c.
Particulate Sampler (RM-A9) 1 31 d.
Effluent System Flow Rate 1
26 Measuring Device (FR-148) e.
Sampler Flow Rate Monitor 1
26
F S
TABLE 3.21-2 9
RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRLHENTATION
?
M MINIMUM CHANNELS
{
INSTRUMENT OPERABLE APPLICABILITY ACTION 4.
Condenser Vent System a.
Low Range Noble Gas Activity Monitor 2(I) 32 (RM-A5Lo and Suitable Equivalent) 5.
Auxiliary and Fuel Handling Building Ventilation System 4
a.
Noble Gas Activity 1
27
)
RM a
RM 6 b.
Iodine Sampler (RM-A8) or 1
31 (RM-A4 and RM-A6) w i
O E3 c.
Particulate Sampler 1
31 (RM-A8) or (RM-A4 and RM-A6) d.
Effluent System Flow Rate 1
26 i
Measuring Devices j
(FR-151, or FR-149 and FR-150) e.
Sampler Flow Rate Monitor 1
26 NOTE (1): For one of the channels, an operable channel may be defined for purposes of this specification and 4.21.2 only as a suitable equivalent monitoring system capable of being placed in service within one hour. A suitable equivalent system shall include instrumentation with comparable sensitivity and response time to the RM-ASto monitoring channel. When the equivalent monitoring system is in service, indication will be continuously available to the operator, either through indication and alarm in the control room or through communication with a designated individual continuously observing local indication.
o TABLE 3.21-2 TABLE NOTATION
- At all times.
- During waste gas holdup system operation.
- Operability is not required when discharges are positively controlled through the closure of WDG-V47, and RM-A8 and FT-151 are operable.
- At all times during containnent purging.
- At all times when condenser vacuum is established.
ACTION 25 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, the contents of the tank nay be released j
to the environment provided that prior to initiating the release:
l l
1.
At lepst two independent samples of the tank's contents are analyzed in accordance with Table 4.22-2, Item A, and l
2.
At least two technically qualified members of the Unit staff independently verify the release rate calculations and verify the discharge valve lineup, t
3.
The Operations & Maintenance Director, Unit 1, shall approve each release.
Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 26 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway nay continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
l ACTION 27 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirenent, effluent releases via this pathway may continue provided grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and l
these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the channel has been declared inoperable.
(See also 3.5.1, Table 3-5.1, Item C.3.f for RM-A9. )
t 3-103 Amendment No. 72, 78, 104 i
o TABLE 3.21-2 TABLE NOTATION (continued)
ACTION 30 1.
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, a grab sample shall be collected and analyzed for the inoperable gas channel (s) at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. With both channels inoperable, a grab sample shall be collected and analyzed for the inoperable gas channel (s):
(a) at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during degassing operations.
(b) at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during other operations (e.g. Feed and Bleed).
2.
If the inoperable gas channel (s) is not restored to service within 14 days, a Special Report shall be submitted to the Regional Administrator of the NRC Region I Office and a copy to the Director, Office of Inspection and Enforcement within 30 days of declaring the channel (s) inoperable. The report shall describe (a) the cause of the monitor inoperability, (b) action being taken to restore the instrument to service, and (c) action to be taken to present recurrence.
ACTION 31 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, effluent releases via this pathway may continue provided that within four hours af ter the channel has been l
declared inoperable, samples are continuously collected with auxiliary sampling equipment.
ACTION 32 With the nunber of channels OPERABLE less than required by the Minimum Channels OPERABLE requirements, effluent releases via this pathway may continue for up to 28 days, provided that one OPERABLE channel remains in service or is placed in service within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. After 28 days, or if one OPERABLE channel does not remain in service or is not placed in service within 1 h'our, the provisions of 3.0.1 apply.
3-104 (p. 3-105 intentionally blank)
(p. 3-105a has been eliminated)
Amendment No. 72, 78, 103, 104
o 3.22 RADI0 ACTIVE EFFLUENTS 3.22.1 LIQUID EFFLUENTS 3.22.1.1 CONCENTRATION LIMITING CONDITION FOR OPERATION 3.22.1.1 The concentration of radioactive material released at anytime from the unit to unrestricted areas (see Figure 5-4) shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 3 x 10-3 uCi/cc total activity.
APPLICABILITY: At all times ACTION:
a.
With the concentration of radioactive material released from the unit to unrestricted areas exceeding the above limits, immediately restore concentration within the above limits.
b.
If action "a" cannot be met, then be in:
1.
At least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 2.
At least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and 3.
At least COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
BASES This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the unit to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will not result in exposures with (1) the Section II.A design objectives of Appendix I,10 CFR Part 50, to an individual and (2) the limits of 10 CFR Part 20.106 (e) to the population. The concentration limit for noble gases is based upon the assumption the Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
3-106 Amendment No. 72
y RA010 ACTIVE EFFLUENTS 3.22.2 GASEOUS EFFLUENTS 3.22.2.1 DOSE RATE LIMITING CONDITION FOR OPERATIONS 3.22.2.1 The dose rate due to radioactive materials released in gaseous effluents from the site (see Figure 5-3) shall be limited to the following:
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.
For I-131, I-133, tritium and all radionuclides in 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:
a.
With the release rate (s) exceeding the above limits, immediately decrease the release rate to comply with the above limit (s) b.
If action "a" cannot be met, then be in:
1.
At least HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 2.
At least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and 3.
At least COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
BASES The specification is provided to ensure that the release rate at anytime at the site boundary from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B. Table II. These limits provick reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendix B. Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)). For individuals who may at times be within the site boundary, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the exclusion area boundary. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the site boundary to less than or equal to 500 mrem / year to the total body or to less than or equal to 3000 mrem / year to the skin. These release rate limits also restrict, at all times, the corresponding thyroid dose rate above background to an infant via the cow-milk-infant pathway to less than or equal to 1500 mrem / year for the nearest cow to the plant.
3-111 Anendment No. 72 l
i 9
w y
RADI0 ACTIVE EFFLUENTS 3.22.2.4 GASEOUS RADWASTE TREATMENT SYSTEM i
LIMITING CONDITION FOR OPERATION 3.22.2.4 The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION' EXHAUST 3 "
TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASEOUS RADWASTE TREATMENT SYSTEM shall be used to reduce radioactive materials in the gaseous waste prior to their discharge when the monthly projected gaseous l
effluent air doses due to untreated gaseous effluent releases from the unit (see Figure 5-3) would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. The appropriate portions of the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the monthly projected doses due to gaseous effluent s l ', '
releases from the site (see Figure 5-3) would exceed 0.3 mrem to any organ.'
.s t
APPLICABILITY: At all times.
ACTION:
a.
With the GASEOUS RADWASTE TREATMENT SYSTEM and/or the VENTILATION EXHAUST TREATMENT SYSTEM inoperable for more than 31 days or with gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which includes the following information:
1.
Identification of the inoperable equipment or subsystems and the reason for inoperability, 2.
Action (s) taken to restore the inoperable equipment to OPERABLE status, and 3.
A summary description of action (s) taken to prevent a recurrence.
BASES _
The use of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST l
TREATMENT SYSTEM ensures that gaseous effluents are treated as appropriate prior to release to the environment. The appropriate portions of this system i
provide reasonable assurance that the releases of radioactive materials in 3
gaseous effluents will be kept "as low as is reasonably achievable." This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50.
The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the guide set forth in Sections II.B and II.C of Appendix I,10 CFR Part 50, for gaseous effluents.
3-115 Amendment No. 72 i
RADI0 ACTIVE EFFLUENTS 3.22.2.6 WASTE GAS DECAY TANKS LIMITING CONDITION FOR OPERATION 1
3.22.2.6 The quantity of radioactivity contained in each waste gas decay tank shall be limited 1 8800 curies noble gases (considered as Xe-133).
APPLICABILITY: At all times.
ACTION:
a.
With the quantity of radioactive material in any waste gas decay l
tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit.
BASES Restricting the quantity of radioactivity contained in each waste gas decay l
tank provides assurance that in the event of an uncontrolled release of the tanks contents, the resultihg total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 15.7.1, " Waste Gas System Failure."
\\
1 3-117 Amendment No. 72 j
l
RADIOLOGICAL ENVIRONMENTAL MONITORING 3.23.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR'0PERATION 3.23.3 Analysis shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by NRC.
APPLICABILITY: At all times.
ACTION:
a.
With analyses not being performed as required above, report the l
corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report.
BASES The requirements for participation in an Interlaboratory Comparison program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.
3-127 Amendment No. 72
e 3I m
h-k TABLE 4.1-1
!I INSTRUMENT SURVEILLANCE REQUIREMENTS
}U CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS k!
1.
Protection Channel NA M
NA Coincidence Logic 2.
Control Rod Drive NA M
NA (1)
Includes independent testing Trip Breaker of shunt trip and undervoltage trip features.
3.
Power Range Amplifier Df1)
NA (2)
(1)
When reactor power is greater than 15%.
(2)
When above 15% reactor power run a heat balance check once per shift. Heat balance calibration shall be performed whenever heat balance exceeds l,
indicated neutron power by more than two percent.
4.
Power Range Channel S
M M(1)(2)
(1)
When reactor power is greater than 60% verify imbalance using incore instrumentation.
(2)
When above 15% reactor power calculate axial offset upper and lower chambers af ter each startup if not done within the previous seven days.
5.
Intermediate Range Channel S(1)
PS/U NA (1)
,When in service 6.
Source Range Channel S(l)
PS/U NA (1)
When in service 7.
Reactor Coolant Temperature S
M R
Channel
@{
TABLE 4.1-1 (Continued)
B j
CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS Ei 8.
High Reactor Coolant S
M R
Pressure Channel u
9.
Low Reactor Coolant S
M R
Pressure Channel
- 10. Flux-Reactor Coolant Flow S
M R
Comparator
- 11. Reactor Coolant Pressure S
M R
Temperature Comparator
- 12. Pump Flux Comparator S
M R
- 13. High Reactor Building S
M R
Pressure Channel
- 14. High Reactor Injection NA Q
NA Logic Channel u
i
- 15. High Pressure Injection Analog Channels a.
Reactor Coolant S(1)
M R
(1)
When reactor coolant system is Pressure Channel pressurized above 300 psig or Tav is greater than 200"F 16.
Low Pressure Injection NA Q
NA Logic Channel 17.
Low Pressure Injection NA Q
NA Analog Channels a.
Reactor Coolant S(1)
M R
(1)
When reactor coolant system is Pressure Channel pressurized above 300 psig or Tav is greater than 200*F
- 18. Reactor Building Emergency NA Q
NA Cooling and Isolation System Logic Channel s
F E
o.
5 TABLE 4.1-1 (Continued) 5 CHANNEL DESCRIPTION CHECK TEST CALIBRATE REMARKS 49.
Saturation Margin Monitor S(l)
M(1)
R (1) When Tave is greater than 525'F.
h 50.
Emergency Feedwater Flow NA M(1)
R (1) When Tave is greater than 250*F.
Instrumentation 51.
Emergency Feedwater Initiation a.
Loss of RCPs NA Q(1)(2)
R (1) When Tave is greater than 250*F.
b.
Loss of Both Feedwater Pumps NA Q(1)(2)
R (2) Includes logic test only.
52.
Backup Incore Thermocouple M(1)
NA R
(1) When Tave is greater than 250*F.
Display
?
O e
J
4.21 RADIOACTIVE EFFLUENT INSTRlNENTATION 4.21.1 RADIOACTIVE LIQUID EFFLUENT INSTRlNENTATION SURVEILLANCE REQUIREMENTS 4.21.1 Each radioactive liquid effluent monitoring instrunentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL TEST operations during the MODES and at the frequencies shown in Table 4.21-1, i
4-87 Amendment No. 72
4.21.2 RADI0 ACTIVE GASE0US PROCESS AND EFFLUENT MONITORING INSTRLMENTATION SURVEILLANCE REQUIREMENTS 4.21.2 Each radioactive gaseous process or effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL TEST operations at the frequencies shown in Table 4.21-2.
l 4-90 Amendnent No. 72
p TABLE 4.21-2 to h
RADI0 ACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
$e g
CHANNEL SOURCE CHANNEL CHANNEL INSTRUMENT CHECK,
CHECK CALIBRATION TEST APPLICABILITY M
1.
Waste Gas Holdup System S
a.
Noble Gas Activity Monitor P
P E(3)
Q(1)
(RM-A7) b.
Effluent System Flow Rate P
N/A E
Q Measuring Device (FT-123) 2.
Waste Gas Holdup System Explosive Gas Monitoring System a.
Hydrogen Monitor D
N/A Q(4)
M b.
Oxygen Monitor D
N/A Q(5)
M 3.
Containment Purge Vent System a.
Noble Gas Activity Monitor D
P E(3)
M(1)
(RM-A9 )
b.
Iodine Sampler (RM-A9)
W N/A N/A N/A c.
Particulate Sampler (RM-A9)
W N/A N/A N/A d.
Effluent System Flow Rate D
N/A E
Q Measuring Device (FR-148) e.
Sampler Flow Rate Monitor D
N/A E
N/A 4.
Condenser Vent System a.
Noble Gas Activity Monitor D
M E(3)
Q(2)
(RM-A5 and Suitable Equivalent -
See Table 3.21-2, Item 4.a)
N
![
TABLE 4.21-2 (Continued) 2 5
RADI0 ACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
.if CHANNEL SOURCE CHANNEL CHANNEL y
INSTRUMENT CHECK CHECK CALIBRATION TEST APPLICABILITY 5.
Auxiliary and Fuel Handling Building Ventilation System a.
Noble Gas Activity Monitor D
M E(3)
Q(1)
(RM-A8) or (RM-A4 and RM-A6) i b.
Iodine Sampler (RM-A8) or (Ri-A4 and RM-A6)
W N/A N/A N/A c.
Particulate Sampler (RM-A8) or (RM-A4 and RM-A6)
W N/A N/A N/A d.
Effluent System Flow Rate
~
Measurement Devices D
N/A E
Q (FR-151, or FR-149 and FR-150) 8 e.
Sampler Flow Rate Monitor D
N/A E
N/A t
l l
TABLE 4.21-2 (Continued)
TABLE NOTATION At all times.
During waste gas holdup system operation Operability is not required when discharges are positively controlled through the closure of WDG-V47, and RM-A8 and FT-151 are operable.
At all times during containment purging At all times when condenser vacuum is established (1)
The CHANNEL TEST shall also demonstrate that automatic isolation of this pathway for the Auxiliary and Fuel Handling Building Ventilation System, the supply ventilation is isolated and control room alarm annunciation occurs if the following condition exists:
1.
Instrument indicates measured levels above the high alarm / trip setpoi nt.
(Includes circuit failure) 2.
Instrument indicates a down scale failure.
(Alarm function only)
(Includes circuit failure) 3.
Instrument controls moved from the operate mode.
(Alarm function only)
(2)
The CHANNEL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:
1.
Instrument indicates measured levels above the alarm setpoint.
(includes circuit failure) 2.
Instrument indicates a down scale failure (Includes circuit failure) 3.
Instrument controls noved from the operate mode.
(3)
The initial CHANNEL CALIBRATION for radioactivity measurement instrumentation shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards should permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration should be used.
(Operating plants may substitute previously established calibration procedures for this requirement. )
(4)
The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
1.
One volume percent hydrogen, balance nitrogen, and 2.
Four volume percent hydrogen, balance nitrogen.
t 4-93 Amendment No. 72
TABLE 4.21-2 (Continued)
TABLE NOTATION (5)
The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
1.
One volume percent oxygen, balance nitrogen, and 2.
Four volume percent oxygen, balance nitrogen.
l 4-94 Amendment No. 72
4.22 RADIOACTIVE EFFLUENTS 4.22.1 LIQUID EFFLUENTS SURVEILLANCE REQUIREMENTS 4.22.1.1 CONCENTRATION 4.22.1.1.A The radioactivity content of each batch of radioactive liquid waste l
shall be determined prior to release by sampling and analysis in accordance with Table 4.22-1.
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 3.22.1.1.
4.22.1.1.B Post-release analysis of samples composited from batch releases shall l
be performed in accordance with Table 4.22-1.
The results of the previous post-release analysis shall be used with the calculational methods in the 0D04 to assure that the concentrations at the point of release were maintaired within the limits of Specification 3.22.1.1.
4.22.1.1.C The radioactivity concentration of liquids discharged from continuous l
release points shall be determined by collection and analysis of samples in accordance with Table 4.22-1.
The results of the analysis shall be used with the calculational methods of the ODCM to assure that the concentration at the point of release is maintained within the limits of Specification 3.22.1.1.
4-95 Amendment No. 72
F
!.l TABLE 4.22-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM 2
l Sampli ng i
Mi nimum l
l Lower Limit Liquid Release Type l
Frequency l
Analysis l
Type of Activity I
of Detection l
l Frequency I
Analysis I
(LLD) l l
l l
(uCi/ml) a I
I I
I l
P l
P l
H-3 l 1 x 10-5 i
A.1 Batch Waste-l Each Batch l
Each Datch l
Principal Gamma 1 5 x 10-7 Release Tanksd l
l l
Emitters f I
i i
l l
l I
l l
I-1 31 l 1 x 10-6 I
I I
l l
l l
l l
P l
l l 1 x 10-4 I
One Batch /M l
M l
Dissolved and l
l l
l Entrained Gases l
l l
l (Gamma Emitters) l I
I I
I l
l I
l a
E I
I I
I.
l P
l Q
l Gross alpha l 1 x 10-7 l
Each Batch l
Composite b l
l l
l l
l Sr-89, Sr-90 l 5 x 10-8 I
I I
I l
l l Fe-55 l 1 x 10-6 I
I I
I
SI k
TABLE 4.22-1 (Continued) fo
?,
RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM
.i!F l
Sampling l
Minimum l
l Lower Limit y."
Liquid Release Type l
Frequency l
Analysis l
Type of Activity 1
of Detection l
l Frequency l
Analysis I
(LLD) co l
l l
l (uCi/ml) a l
I W
I I
A.2 Continuous e I
Continuous c l
Compositec l
Principal Gamma l 5 x 10-7 Releases l
I l
Emitters f l
I I
I I
l l
I I-1 31 l 1 x 10-6 I
I I
I I
I I
I I
M i
I i 1 x 10-5 l
M I
Dissolved and l
l l
1 Entrained Gases i
l l
I (Gamma Emitters) l I
I I
I I
I I
I I
I I
I f
l I
M l
H-3 1 1 x 10-5 I
Continuous c l
Composite c l
l l
l l
Gross alpha l 1 x 10-7 1
I I
l l
I I
I l
l Q
l Sr-89, Sr-90 l 5 x 10-8 l
Continuous c I
composite c l
l 1
l l
l Fe-55 l 1 x 10-6 I
I I
I I
I I
I
Table 4.22-1 (Continued)
TABLE NOTATION a.
The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system (which may include radiochemical separation):
4' 8b LLD =
E x V x 2.22 x 106 x Y x exp (-AA t)
LLD is the "a priori" lower limit of detection as defined above (as microcurie per unit mass or volume),
sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),
E is the counting efficiency (as counts per transformation),
Y is tl.e sample size (in units of mass or volume),
2.22 x 106 is the number of transformations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),
A is the radioactive decay constant for the particular radionuclide, and A t is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).
The value of sb used in the calculation of LLD for a detection system shall be based on the actual observed variance of the background coun-ting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. Typical values of E, V, Y, and A t shall be used in the calculation.
4 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 represen-tative of the liquids released.
4-98 Amendment No. 72
c.
To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream.
Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.
d.
A batch release is the discharge of liquid wastes of a discrete volume.
Prior to sampling for analyses, each batch shall be isolated, and the thoroughly mixed, by a method described in the ODCM, to assure representative sampling.
e.
A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g., from a volume or system that has an input flow during the continuous release.
f.
The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measureable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses shall be reported as "less than" the nuclide's LLD, and shall not be reported as being present at the LLD level for that nuclide. The "less than" values shall not be used in the required dose calculations.
4-99 Amendment No. 72
4.22.1.2 DOSE CALCULATIONS Cumulative dose contributions from liquid effluents shall be determined in accordance with the Offsite Dose Calculation Manual (ODCM) at least once a month.
4.22.1.3 LIQUID WASTE TREATMENT l
4.22.1.3.1 Doses due to liquid releases shall be projected at least once a month, in accordance with the ODCM.
4.22.1.3.2 The liquid radwaste treatment system shall be demonstrated OPERABLE by operating the liquid radwaste treatment system equipment for at least 60 minutes quarterly unless the liquid radwaste system has been utilized to process radioactive liquid effluents during the previous 92 days.
4.22.1.4 LIQUID HOLDUP TANKS The quantity of radioactive material contained in each of the tanks specified in Specification 3.22.1.4 shall be determined to be within the limit by analyzing a representative sample of the tank's content l
weekly when radioactive materials are being added to the tank.
l 1
4-100 Amendment No. 72
4.22.2 GASEOUS EFFLUENTS SURVEILLANCE REQUIREMENTS 4.22.2.1 DOSE RATES 4.22.2.1.1 The dose rate due to noble gases in gaseous effluents shall be determined to be within the limits of Specification 3.22.2.1.a in accordance with the methods and procedures of the ODCM.
4.22.2.1.2 The dose rate of radioactive materials, other than noble gases, in gaseous effluents shall be determined to be within the limits of Specification 3.22.2.1.b in accordance with methods and piocedures of l
the ODCM by obtaining representative samples and perfcrming analyses in accordance with the sampling and analysis program, specified in Table 4.22-2.
4 -1 01 Amendment No. 72
___,_____-r 4,-
.._ _.. _y
TABLE 4.22-2 y
RADI0 ACTIVE GASE0US WASTE SAMPLING AND ANALYSIS PROGRAM E
l Sampli ng l
Minimum l
I Lower Limit U
Gaseous Release Type l
Frequency l
Analysis i
Type of Activity I
of Detection l
l Frequency l
Analysis I
(LLD) l l
l l
(uCi/ml) a l
l l
l 1
P I
P l
I A.
Waste Gas l
Each Tank l
Each Tank l
Principal Gamma l 1 x 10-4 Decay Tank l
Grab l
l Emitters 9 l
l Sample l
1 l
l l
l l
l l
1 I
B.
Containment l
Pb l
pb l
l j x 10-4 Purge l
Each Purge l
Each Purge l
Principal Gamma l
l Grab l
l Emmitters 9 l
l Sample l
l H-3 l 1 x 10-6 l
l l
l l
l l
1 C.
Auxiliary and l
l l H-3 l 1 x 10-6
[
Fuel Handling l
M,c,e l
M l
Principal Gamma l 1 x 10-4 Buil ding l
Grab l
l Emitters 9 l
o Ventilation l
Sample l
l l
l l
l l
1 I
I I
I D.
All Release l
Continuous f l
Wd l
I-131 l 1 x 10-12 Type as l
l Charcoal l
l listed in A, B, I
l Sample l
l C above.
l l
l l
l l
l l
l l
Wd l
Principal Gamma i 1 x 10-11 l
Continuous f l
Particulate l
Emitters 9 l
l l
l (I-131, Others) l l
l 1
l l
1 I
I l
l Q
l l 1 x 10-11 l
Continuous f l
Composite i
gross alpha l
l l
Particulate l
I Sample
5"
~
TABLE 4.22-2 RADI0 ACTIVE GASE0US WASTE SAMPLING AND ANALYSIS PROGRAM
' E l
Sampling l
Minimum l
l Lower Limit M
Gaseous Release Type l
Frequency l
Analysis l
Type of Activity I
of Detection l
l Frequency l
Analysis I
(LLD) l l
l l
(uCi/ml)a I
I I
I I
I I
I D.
All Release l
Continuous f l
Q l
l
~
Type as l
l Composite l
Sr-89, Sr-90 l 1 x 10-11 listed in A, B, I
l Particulate l
l C above.
I l
Sample l
l l
l l
l l
l l
l E.
Condenser vacuum l
M, h l
l l
Pumps Exhaust 9 l
M, h l
Principal Gamma l 1 x 10-4 l
l l
Emitters 9 l
l l
l H-3 l 1 x 10-6 p
i l
i I
5 i w P
Table A.22-2 (Continued)
TABLE NOTATION a.
The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system (which may include radiochemical separation):
4.66 sb RD=
E x V x 2.22 x 106 x Y x exp (- AAt)
Where:
LLD is the "a priori" lower limit of detection as defined above (as microcurie per unit mass or volume),
sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),
E is the counting efficiency (as counts per transformation),
V is the sample size (in units of mass or volume),
2.22 x 106 is the number of transformations per minute per microcurie, Y is the fractional radiochemical yield (when applicable),
A is the radioactive decay constant for the particular radionuclide, and A t is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).
The value of sb used in the calculation of LLD for a detection system shall be based on the actual observed variance of the background coun-ting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. Typical values of E, V, Y, and At shall be used in the calculation.
b.
Applicable only when condenser vacuum is established.
Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within one hour unless (1) ar.41ysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.
c.
Tritium grab samples shall be taken at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the refueling canal is flooded.
4-104 Amendment No. 72
l l
Table 4.22-2 (Continued)
TABLE NOTATION d.
Charcoal cartridges and particulate filters shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after changing (or af ter removal from sampler).
e.
Tritium grab samples shall be taken weekly from the spent fuel pool area whenever spent fuel is in the spent fuel pool, f.
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 3.22.2.1, 3.22.2.2, and 3.22.2.3.
g.
The principal ganna emitters for which the LLD specification applies exclusively are the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135 and Xe-138 for gaseous emmissions and Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, 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 measureable and identifiable, together with the above nuclides, shall also be identified and reported.
Nuclides which are below the LLD for the analyses shall be reported as "less than" the nuclide's LLD, and shall not be reported as being present at the LLD level for that nucli de.
The "less than" values shall not be used in the required dose calculations, h.
Applicable only when condenser vacuum is established.
Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15 percent of RATED THERMAL POWER within one hour unless (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.
l 4-105 Amendment No. 72
\\
,s 4.22.2.2 DOSE, NOBLE GAS Cumulative dose contributions from noble gas effluents for the current calendar quarter and current calendar year shall be determined in accordance with the OFFSITE DOSE CALCULATION MANL'AL (0D01) monthly.
4.22.2.3 DOSE, RADI0 ACTIVE MATERIAL IN PARTICULATE FORM, RADIONUCLIDES OTHER THAN NOBLE GASES Cumulative dose contributions from radioactive material in particulates, radiofodines and radionuclides other than noble gases for the current calendar quarter and current calendar year shall be determined in accordance with the OFFSITE DOSE CALCULATION MANUAL (0DC4) monthly.
4.22.2.4 GASEOUS WASTE TREATMENT Doses due to gaseous releases from the unit shall be projected monthly in accordance with the ODCM.
4.22.2.5 EXPLOSIVE GAS MIXTURE The concentrations of hydrogen and oxygen in the waste gas holdup system shall be determined to be within the limits of TS 3.22.2.5 by monitoring the waste gases in the Waste Gas Holdup System with the hydregen and oxygen monitors covered in Table 3.21-2 of Specification 3.21.2.
4.22.2.6 WASTE GAS DECAY TANK The concentration of radioactivity contained in the vent header shall be determined weekly.
If the concentration of the vant header exceeds 10.7 Ci/ce, daily samples shall be taken of each waste gas decay tank being added to, to determine if the tank (s) is < 8800 Ci/ tank.
4-106 Amendment No. 72
.+
4.22.3 SOLID RADI0 ACTIVE WASTE SURVEILLANCE REQUIREMENTS 4.22.3.1 SOLID RADWASTE SYSTEM The solid radwaste system shall be demonstrated OPERABLE quarterly by:
a.
Operating the solid radwaste system at least once in the previous 92 days in accordance with the PROCESS CONTROL PROGRAM or; b.
Verification of the existence of a valid contract for SOLIDIFICATION to be performed by a Contractor in accordance with a PROCESS CONTROL PROGRAM.
4.22.3.2 PROCESS CONTROL PROGRAM The PROCESS CONTROL PROGRAM shall be used to verify the SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of radioactive waste required to be solidified by the Process Control Program.
a.
If any test specimen fails to verify SOLIDIFICATION, the 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 Process Control Program, and a subsequent test verifies solidification.
Solidification of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM.
b.
If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until 3 consecutive initial test specimens demonstrate SOLIDIFICATION.
The PROCESS CONTROL PROGRAM shall be modified as required, to assure SOLIDIFICATION of subsequent batches of waste.
4-107 Amendment No. 72
e...
i 4.22.4 TOTAL DOSE SURVEILLANCE REQUIREMENT i
4.22.4.1 DOSE CALCULATION Cumulative dose contributions from liquid and gaseous effluents shall be ~ determined in accordance with TS 4.22.1.2, 4.22.2.2 and 4.22.2.3 and in accordance with the ODCM.
4 4
4-108 Amendment No. 72
D4 g e 4.23 RADIOLOGICAL ENVIR0fMENTAL MONITORING SURVEILLANCE REQUIREMENTS 4.23.1 MONITORING PROGRAM The radiological environmental monitoring samples shall be collected pursuant to Table 3.23-1 from the locations given in the tables and figures in the ODCM and shall be analyzed pursuant to the requirements of Table 3.23-1 and 4.23-1.
4.23.2 LAND USE CENSUS The land use census shall be conducted at least once per 12 months. The information may be obtained through the use of door-to-door survey, aerial survey, or consjtation with local agriculture authorities.
4.23.3 INTERLABORATORY COMPARISON PROGRAM A summary of the results obtained as part of the Inter-Laboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report.
l 4-1 09 i
j Amendment No. 72 t
1
[
1
'l
[
h TABLE 4.23-1 MAXIMlM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)a,c z
Airborne Particulate w
Water or Gas Fish Milk Food Products Sediment Analysis (pCi/1)
(pCi/m3)
(pCi/Kg, wet)
(pCi/1)
(pCi/Kg, wet)
(pCi/Kg, dry) gross beta 4
1 x 10-2 3H 2000 54Mn 15 130 59Fe 30 260 58,60 o 15 130 C
65Zn 30 260 95Zr 30 95Nb 15 1 31 1 lb 7 x 10-2 1
60 l
i 1 34 s 15 5.x 10-2 130 15 60 150 C
137 s 18 6 x 10-2 150 18 80 180 l
C 140 a 60 60 B
140 a 15 15 L
o Table 4.23-1 (Continued)
TABLE NOTATION a.
The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.
For a particular measurement system (which may include radiochemical separation):
4*
8b LLD =
E x V x 2.22 x Y x exp (- AAt)
Where:
LLD is the "a priori" lower limit of detection as defined above (as picocurie per unit mass or volume),
sb is the standard deviation of the background counting rate or of 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),
2.22 is the number of transformations per minute per picocurie, Y is the fractional radiochemical yield (when applicable),
A is the radioactive decay constant for the particular radionuclide, and A t for environmental samples is the elapsed time between sample collection, or the end of the sample collection period, and time of counting.
The value of sb used in the calculation of LLD for a detection system shall be based on the actual observed variance of the background coun-ting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. Typical values of E, V, Y, and At shall be used in the calculation, b.
LLD for drinking water.
c.
Other peaks which are measured and identifiable, together with the radioactivity in Table 4.23-1, shall be identified and reported.
4 -111 Amendment No. 72
--