ML18011A616
| ML18011A616 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 10/24/1994 |
| From: | CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML18011A615 | List: |
| References | |
| RTR-NUREG-1431 NUDOCS 9411020022 | |
| Download: ML18011A616 (48) | |
Text
ENCLOSURE TO SERIAL: HNP-94-073 ENCLOSURE 1 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT RELOCATION OF SELECTED TECHNICAL SPECIFICATIONS IN ACCORDANCE WITH NUREG-1431 BASIS FOR CHANGE RE UEST
~Back round The Commission's Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors (58 FR 39132), dated July 22, 1993, provides specific criteria for the content of Technical Specifications (TS). The Final Policy Statement specifically recognizes that:
"The purpose of Technical Specifications is to impose those conditions or limitations upon reactor operation necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety by identifying those features that are of controlling importance to safety and establishing on them certain conditions of operation which cannot be changed without prior Commission approval."
The Final Policy Statement references NUREG-1431, "Standard Technical Specifications, Westinghouse Plants" and encourages licensees to implement a program to upgrade their TSs consistent with the purpose stated above. In addition, the Commission has also indicated a willingness to entertain requests to adopt portions of the improved Standard Technical Specifications (STS) as line-item improvements.
The Final Policy Statement further describes four criteria to delineate those constraints on design and operation of nuclear power plants that are derived from the plant safety analysis report or probabilistic safety assessment (PSA) information and that belong in TS in accordance with 10 CFR 50.36 and the purpose of TS stated above. Therefore, limiting conditions for operations (LCOs) which do not meet any of these four criteria may be proposed for removal from the TS and relocated to licensee-controlled documents such as the Final Safety Analysis Report (FSAR).
Pro osed Chan e This amendment proposes to revise Technical Specifications 3.3.4, "Turbine Overspeed Protection," 3.7.12, "Area Temperature Monitoring," and 3.11.2.6, "Gas Storage Tanks," and associated bases, consistent with NUREG-1431, the new Standard Technical Specifications for Westinghouse Plants. The proposed revisions are also consistent with the Commission's Final Policy Statement for relocation of current Technical Specifications.
9411020022 941024 El-1 PpDR ADOCK 05000400 page PDR
ENCLOSURE TO SERIAL'NP-94-073 Basis The Commission's Final Policy Statement states that TS that do not meet any of the screening criteria for retention may be proposed for removal from TS and relocated to licensee-controlled documents such as the FSAR.
In November 1987, Westinghouse submitted to the NRC, WCAP-11618 which applied the Commissions's screening criteria to the Westinghouse Standard TS (NUREG-0452, Revision 4 and draft Revision 5). The Commission documented the results of their review of WCAP-11618 in a letter dated May 9, 1988 to R. A. Newton, Chairman of the Westinghouse Owners Group.
Shearon Harris Nuclear Power Plant (SHNPP) TS are typical of Westinghouse three-loop plants and are based on NUREG-0452, draft Revision 5. As such, WCAP-11618 applies to the SHNPP TS, except: for those SHNPP LCOs that are not evaluated in WCAP-11618. Table 1 documents application of the screening criteria to those TS addressed by this request. The results are based on the application of the criteria and the NRC review of WCAP-11618.
The Commission's Final Policy Statement states that licensees may submit license amendment requests based on the Final Policy Statement and that licensees should identify the location and administrative controls for the relocated requirements.
The relocated TS will be included in the FSAR Update or appropriate plant procedures.
Table 1
SUMMARY
OF CRITERIA APPLICATION TECHNICAL SPECIFICATION RELOCATED SHNPP STS Rev 5 Technical Specification NRC SHNPP Number Number Title Results Results 3.3.4 3.3.4 Turbine Overspeed Relocate Relocate Protection 3.7.12 3.7.13 Area Temperature Relocate Relocate Monitoring 3.11.2.6 3.11.2.5 Gas Storage Tanks Relocate Relocate Page El-2
ENCLOSURE TO SERIAL: HNP-94-073 Conclusions The TS proposed for relocation in this amendment request do not meet any of the Commission's screening criteria for retention. Consistent with NUREG-1431 and with the Commission's Final Policy Statement for relocation of current Technical Specifications these TS may be relocated into appropriate plant administrative controls and referenced in the FSAR Update.
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ENCLOSURE TO SERIAL: HNP-94-073 ENCLOSURE 2 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT RELOCATION OF SELECTED TECHNICAL SPECIFICATIONS IN ACCORDANCE WITH NUREG-1431 10 CFR 50.92 EVALUATION The Commission has prov'ided standards in 10 CFR 50.92(c) for determining whether a significant hazards consideration exists, A proposed amendment to an operating license for a facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety, Carolina Power 6 Light Company has reviewed this proposed license amendment request and determined that its adoption would not involve a significant hazards determination. The bases for this determination are as follows; Pro osed Chan e This amendment proposes to revise Technical Specifications (TS) 3.3.4, "Turbine Overspeed Protection," 3.7.12, "Area Temperature Monitoring," and 3.11.2.6, "Gas Storage Tanks," and associated bases, consistent with NUREG-1431, the new Standard Technical Specifications for Westinghouse Plants. The proposed revisions are also consistent with the Commission's Final Policy Statement for relocation of current Technical Specifications.
Basis This change does not involve a significant hazards consideration for the following reasons:
The proposed 1
amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.
These proposed changes will simplify the TS, and implement the recommendations of the Commission's Final Policy Statement on TS Improvements. Since the elements of these TS are being relocated to licensee-controlled documents, any future changes would be controlled under 10 CFR 50.59. The proposed changes are administrative in nature and do not involve any modifications to any plant equipment or affect plant operation. Therefore, there would be no increase in the probability or consequences of an accident previously evaluated.
The proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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ENCLOSURE TO SERIAL: HNP-94-073 The proposed changes are administrative in nature, do not involve any physical alterations to plant equipment, and result in no change in the method by which any safety-related system performs its function.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed amendment does not involve a significant reduction in the margin of safety.
These changes do not affect any Final Safety Analysis Report (FSAR)
Chapter 15 accident analyses or have any impact on margin as defined in the Bases to the Technical Specifications. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
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ENCLOSURE TO SERIAL: HNP-94-073 ENCLOSURE 3 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT RELOCATION OF SELECTED TECHNICAL SPECIFICATIONS IN ACCORDANCE WITH NUREG-1431 ENVIRONMENTAL CONSIDERATIONS 10 CFR 51.22(c)(9) provides criterion for and identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; (3) result in a significant increase in individual or cumulative occupational radiation exposure. Carolina Power 6 Light Company has reviewed this request and determined that the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the amendment. The basis for this determination follows:
Pro osed Chan e This amendment proposes to revise Technical Specifications 3.3.4, "Turbine Overspeed Protection," 3.7.12, "Area Temperature Monitoring," and 3.11.2.6, "Gas Storage Tanks," and associated bases, consistent with NUREG-1431, the new Standard Technical Specifications for Westinghouse Plants. The proposed revisions are also consistent with the Commission's Final Policy Statement for relocation of current Technical Specifications.
Basis The change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the following reasons:
As demonstrated in Enclosure 2, the proposed amendment does not involve a significant hazards consideration.
- 2. The proposed amendment does not result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite.
The proposed changes are administrative in nature and do not iqvolve any new equipment,, or require existing systems to perform a different type of function than they are currently designed to perform. The change does not introduce any new effluents or increase the quantities of existing effluents. As such, the change can not affect the types or amounts of any Page E3-1
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ENCLOSURE TO SERIAL: HNP-94-073 effluents that may be released offsite.
The proposed amendment does not result in a significant increase in individual or cumulative occupational radiation exposure.
The proposed changes are administrative in nature and do not result in any physical plant changes or new surveillance. The relocation of TS to licensee-controlled documents can have no impact on radiation exposure since the elements of these TS will be retained and controlled under 10 CFR 50.59. Therefore, the amendment has no affect on either individual or cumulative occupational radiation exposure.
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ENCLOSURE TO SERIAL: HNP-94-073 ENCLOSURE 4 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-.63 REQUEST FOR LICENSE AMENDMENT RELOCATION OF SELECTED TECHNICAL SPECIFICATIONS IN ACCORDANCE WITH NUREG-1431 PAGE CHANGE INSTRUCTIONS Removed Pa e Inserted Pa e vi vi xii xii xiii xiii XV 3/4 3-89 3/4 3-89 3/4 7-28 3/4 7-28 3/4 7-29 3/4 7-29 3/4 11-16 3/4 11-16 B 3/4 3-6 B 3/4 3-6 B 3/4 7-5 B 3/4 7-5 B 3/4 11-5 B 3/4 11-5
ENCLOSURE TO SERIAL: HNP-94-073 ENCLOSURE 5 SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT RELOCATION OF SELECTED TECHNICAL SPECIFICATIONS IN ACCORDANCE WITH NUREG-1431 TECHNICAL SPECIFICATION PAGES
INDEX LIMITINC CONDITIONS FOR OPERATION 'AND SURVElLLANCE REQUIREMENTS SECTION PACE
~ TABt.E 3-3-6 RADIATION MONITOR INC INSTRUMENTATIOH FOR PLANT OPERATIOHSo o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
REQUIREMENTS'~ ~............. 3/4 3-51 TABLE 4.3-3 RADIAT10N MONITORIHC INSTRUMENTAT10N FOR PLANT OPERATIOHS SURVEILLANCE .. ~ ~ ~ .. ~ ........ ~ . 3/4 3-54 Movable lncore Detectors................................. 3/4 3-56 Seismic Instrumentation...............................,.. 3/4 3-57 TABLK 3.3"7 SEISMIC HONITORINC INSTRUMENTATIOH.................... 3/4 3"58 TABLE 4.3-4 SEISMlC MONITORIHC INSTRUHENTATION SURVEIt.LANCE REQUIREMENTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-59 Meteorological. Instrumentation........................... 3/4 3-60 TABLE 3.3-8 METEOROLOCICAL MOHITORIHC INSTRUMEHTATIOH...~ ......... 3/4'-61 TAStE 4.3-5 METEOROLOCICAL MOHZTORINC INSTRUMENTATION SURVEILLANCE REQUIREMENTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ oo ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ o ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3"62 Remote Shutdown System................................... 3/4 3"63 TABLE 3.3-9 REMOTE SHUTDOWN SYSTEMIC ~ .. .... ~ ~ ~ ........ . .....'.
~ ~ ~ 3/4 3-64 TABLE 4.3-6 REMOTE SHUTDOWN MONITORINC INSTRUHENTATIOH SURVEILLANCE REQUIREMENTS ~ ~ ~ . ~ . ~ . ~ .. ~ . ~... ~... ~...... 3/4 3-65 Accident Mani tor ing Ins t rumentat ion.................... 3/4 3-66 ACCIDENT MONITORIHC INSTRUMENTATIONo ~....... ~......
REQUIREMENTS' TABtE 3.3-10 3/4 3-68 TABLE 4.3-7 ACCIDENT MONITORINC INSTRUMEHTATION SURVEILLANCE
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o'/4 3-70 TABLE 3.3-11 (DELETED) ...o..
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~ ~ ~ 3/4 3-73 Hetal Impact Monitoring System........................... 3/4 3-74 Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-75 TABLE 3 ~ 3-12 RADIOACTIVE LIQUZD KFFLUEHT MOHZTORINC INSTRUHENTATIONo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-76 TABLE 4 '-8 RADIOACTIVK LIQUID EFFLUENT MONZTORIHC INSTRUMENTATIOH SURVEILLANCE REQUZREMENTSo ~ ~ ~ ~ .~~~~~~~~ 3/4 3-79 Radioactive Gaseous Effluent Monitoring Instrumentation 3/4 3-82 TABLE 3.3-13 RADIOACTIVE CASKOUS EFFLUENT MONITORINC INSTRUHENTATIONo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ o ~ 3/4 3-&3 TABLE 4 ~ 3-9 RADIOACTIVE GASEOUS EFFLUENT MONITORING
, INSTRUMENTATION SURVEILLANCE QUIREMENTS ~.... ~ ~ .. ~... 3/4 3"86 3/4.3.4 TUR HE OVE PEED P OTECTIOH
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INDEX LIMITING CONDITIONS FOR OPERATXON AND SURVEILLANCE RE UIREMENTS BECTTOC PAGE 3/4 ~ 7 ~ 2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION.~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-10 3/4 7 ~ 3 COMPONENT COOLING WATER SYSTEM. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-11 3/F 7.4 EMERGENCY SERVICE MATER SYSTEM .~~o~~~~ .~~~~~ . ~ ~ ~ ......., 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK....;......... .... .. ...........,.... ~ ~ ~ 3/4 7-t3 3/4.7.6 CONTROL ROOM EMERGENCY FILTRATION SYSTEM ...... ,...,..., ~ 3/4 7-14 3/4.7.7 REACTOR AUXILIARY BUXLDIHG (RAB) EMERGENCY EXHAUST SYSTEMo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7"17 3/4o7 ~ 8 SHUBBERS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-ae FIGURE 4o7 1 (DELETED) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~~ ~ ~ ~ ~~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-24 3/4o 7 ~ 9 SEALED SOURCE CONTAMINATXOMo~ ~ ~ ~ ~ ~ ~ ~~ ~ ~~ ~ ~ ~ ~~~ ~ ~ ~~ ~ ~ ~ ~ ~ ~ ~ 3/4 7"25 3/4o7ol0 (DELETED) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~ ~ ~~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-27 TABLE 3 ~ 7 3 ( DELETED ) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-27 TABLE F 7-4 (DELETED) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-27
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3/4 3/4 3/4 3/4 3/4 7-27 7-27 7-28 7-29 7-30 3/4.8 ELECTRICAL POWER SYSTEMS 3/4 8ol A C. SOURCES Operatxagoo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 8-1 TABLE 4 ~ 8-1 DIESEL GENERATOR TEST SCHEDULEo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ 3/4 8-10 Shatciolc o ~ ~ ~ ~ ~ ~ ~ ~~~~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~ ~ ~ ~ ~~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ o o ~ ~ 3/4 8-11 3/4 8 2 DoC SOURCES 0 PeratXagoooooi ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ 3/4 8-12 TABLE 4o8-2 MITERY SURVEILLANCE REQUIREMENTS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ . 3/4 8"14 Shatdeco ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 8-15 3/4.8 3 OHSITE POWER DZSTRIBUTZON Operatxago ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ .~~~~~~~~~~~~~~~~~~~ . ~ .... 3/4 8-16 Shatdolco ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4,8-18 SHEARON HARRIS - UHZT 1 haaadment No. ~ ~
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INDEC LIMITING CONDITIONS FOR OPERATION ANO SURVEILLANCE REOUIR MENTS SiCTIQN PAGE 3/4. 11 RADIOACTIVE ErFrLUEHTS 3/4. 11. 1 LIQUID EFFLUEHTS C onCNtr atl Ono o ~ ~ ~ ~ o ~ ~ o o o ~ o o ~ o e o o ~ o o e
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~ ~ ~ ~ 3/4 11-1 TABLE 4.11-1 RADIOACTIVE LI(UID VASTE SAMPLIHG AHO ANALYSIS ROGRAMo ~ ~ ~ ~ ~ ~ ~ ~ o o ~ o ~ ~ ~ ~ o ~ o o o ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ e o ~ ~ ~ e o o ~ ~ ~ ~ ~ 3/4 11-2 0 OSeo ~ ~ ~ ~ o ~ ~ o ~ o o ~ ~ o ~ o ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ o ~ e ~ o o ~ ~ ~ ~ o ~ ~ ~ ~ o ~ ~ ~ e ~ o ~ ~ ~ 3/4 11-9 Liquid Radwaste Treatment System......................... 3/4 11-6 Liquid Holdup Tanks...................................... 3/4 11"7 3/4. 11. 2 GASFQUS EFFLUEHTS 0 ose Rate........ 3/4 11-8 TABLE 4o11-2 RADIOACTIVE GASFQUS HASTE SAMPLIHG ANO ANALYSIS P ROGRAMo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ o o ~ o ~ ~ ~ ~ o o e ~ ~ ~ ~ ~ ~ o 3/4 11-9 0 osee Hob1e Gases....................................... 3/4 11-12 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive.
Mataria1 in Particu1ate Form............................. 3/4 11-13 Gaseous Radwasta Treatwt System............'.......;.... 3/4 11-14 Exp 1 os'e Gas M'f puree ~ o ~ ~ o ~ ~ o o ~ o o o o ~ ~ o e o ~ o ~ o ~ ~ ~ ~ o ~ ~ ~ ~ ~ o o 3/4 11-15 (Se4e o e a ~ o ~ ~ ~ e o e e o ~ e o e e o ~ oo e ~ o e ~ o o e e ~ e e e e e o oe o e 3/4 11-16 [
3/4. 11. 3 -SOLID RADIOACTIVE MARo e o o o o o ~ e o o o o ooo o o o ~ o o o ~ oo oo ~ ~ 3/4 11-17 3/4. 11. 4 TOTAL OOSEo ~ e ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ e ~ o ~ ~ ~ ~ ~ ~ ~ ooe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ 3/4 11" 19 3/4. 12 RADIOLOGICAL EHVIRQHMEHTAL MONITORING 3/4o 12o 1 MONITORIHG PROGRAMo ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 I2-1 TABLE 3. 12-1 RAGIOLQGICAI EHVIRQNMEHTAL MONITORING PROGRAM..... 3/4 12-3 TABLE 3.12-2 REPORTXHG LEVELS FOR RAQIOACTIVITY CONCENTRATIONS IN EHVIROOEHTAL SAMPLESo ~ o ~ ~ ~ o ~ ~ o ~ o e ~ o o e o ~ ~ o ~ o ~ o o ~ o o ~ 3/4 12-9 TABLE 4.12-3. DETECTION CAPABILITIES FOR ENVIRONMENTAL.SAMPLE AHALYSISe ~ ~ o ~ ~ ~ e ~ oooo ~ ~ ~ ~ ~ oo ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ oo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 12 1Q 3/4o 12o 2 LAHO USE CENSUS o ~ ~ ~ o ~ ~ ~ ~ ~ o ~ ~ \~ e ~ e ~ ~ ~ e ~ o ~ o e ~ o e ~ ~ ~ ~ ~ ~ ~ ~ \~~\ 3/4 12-13 3/4. 12. 3 IHTERLABORATORY COMPARISON PROGRAM....................... 3/4 12-14 SHEARON HARRIS -'UNIT 1
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'/4 3/4.1 ' HOVABLE CONTROL ASSEMBLIES ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . ~ B 3/a 1-3 3/4 ' POWER DISTRIBUTION LIMITS' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 2-1 3/4 ~ 2 ~ 1 AXIAL PLUX DIFFERENCE>> ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ >> ~ B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, AHD RCS FLOM RATE AND NUCLEAR EHTlQLPY RISE HOT CHANNEL PACTOR>> ~ ~ ~ ~ ~ ~ ~ B 3/4 2-2 FZCURE B 3/4.2-1 ( DELETED) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 2-3 3/4>>2 ' qUADRANT HER TILT RATIO>> ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 2-5 3/4>>2 ' DMB PARAHETERS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 2-6 3/4.3 ZNSTRUHENTATIOH 3/4.3.1 and 3/4 ~ 3.2 REACTOR TRIP SYSTEH AHD ENGINEERED SAPETY FEATURES ACTUATIOM SYSTEH IHSTRUHEHTATZON ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 3-1 3/4.3.3 HOHITORIHC INSTR ATIOM>> ~ ~ ~ ~....... ~ ~ ~ ~ B 3/a 3-3 3/4 ~ 3.4 'HE VERS ED OTE 0 ~ ~ ~ ~ ~ B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEH 3/a.a.l REACTOR COOLANT LOOPS AHD COOLANT CIRCULATZOH>> ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ B 3/4 4-1 3/4.4.2 SAFETY VALVES~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 4-1 3/4 ' ' PRES 8URI Z ER>> ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 4-2 3/4.4.4 RELIEF VALVES ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 a-2 3/a.a.s STEAH CKRATORS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEH LME o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ B 3/4 4-3 3/4.4 7 CHEHISTRY>> ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY>>~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4'4-s 3/4.4.9 PRESSURE/TEHPERATURE LIHITS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ o ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ B 3/4 4-6 SHEARON HARRIS - UHIT 1 Amendment No.~
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INOEX BASES SECTION PAGE 3/4. 7 PLANT SYSTEMS 3/4,7.1 TURBINE CYCLE..........----.-..---- 8 3/4 7"1 3/4.7.2 ~i GENERATOR PRESSURE/T&r. PERATURE LIMITATION........... 8 3/4 7-Z 3/4.7.3'OMPONENT COOLING MATER SYSTEM/. 8 3/4 7-3 3/4.7.4 PERGEHCY SERVICE MATER SY~.....'........... 8 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK.. 8 3/4 7"3 3/4.7. 6 CONTROL ROOM EMERGEHCY FILTRATION SYSTBI.................. 8 3/4 7-3 3/4.7. 7 REACTOR AUXILIARY'UILDIHGBIERGEHCY EXHAUST SY~........ 8 3/4 7"3 3/4. 7. 8 SHUBBERS. 8 3/4 7-4 3/4.7. 9 SEALEQ SOURCE CQNTAMIHATION....... -........ 8 3/4 7-5 3 /4o 7s 10 (DELETED) o ~ ~ ~ ~ o o e ~ ~ ~ o e o e e ~ ~ ~ o o e o ~ o o e o o ~ ~ ~ ~ o o ~ o ~ ~ o ~ ~ ~ o o ~ ~ ~ ~ 8 3/4 7-6 3/4s 7s 11 (DELETED) o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ et o o ~ o ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ o ~ ~ 8 3/4 7-6 3/4.7.12 AR TEtl E ITO HGo ~ ~ os 7opH-87')
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~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ a Sr~ I-E (
3/4 7.13 ESSENTIAL SERVICES CHILLEO MATER SYSTEM................... 8 3/4 7-6 3/4. 8 ELECTRICAL PQMER SYSTEMS 3/4.8.1, 3/4;8.2,.AHO 3/4.8.3 A.C. SOURCES, O.C. SOURCES, AHO QNSITE PQMER DISTRIBUTION....,...................,;...... 8 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES..."................ 8 3/4 8"3 3/4. 9 REFUELIHG OPERATIONS 3/4.9. 1 BORON CONCEHTWION.................-....----.-.--"----.- 8 3/4 9-1 3/4.9.2 IHSTRUMENTATIONo~ ~ e ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ o ~ o ~ e o ~ ~ e ~ ~ ~ ~ 8 3/4 9-1 3/4.9. 3 0 ECAY TIMEo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~
' ~ ~ ~ ~ ~ ~ ~ ~ s 8 3/4 9-1 3/4.9. 4 CQNTAIHM'EHT BUILDING PEHETRATIOHSo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ 8 3/4 9-1 3/4.9. 5 CQ%% NICATIQNSo o ~ o ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ o o o ~ ~ o o ~ ~ ~ o ~ ~ ~ ~ o ~ o o ~ ~ ~ ~ o o 8 3/4 9-1 3/4.9. 6 REFUELIHG MACHIHEo o o o o o o e e o o ~ ~ ~ o ~ ~ ~ e o ~ ~ o ~ o e ~ ~ ~ o ~ o o ~ 8 3/4 9-2 3/4. 9. 7 CRANE TRAVEL - FUEL HANDLIHG BUILDING.............. 8 3/4 9-2 3/4.9. 8 RESIDUAL HEAT REMOVAL AHO COOLANT CIRCULATION...... 8 3/4 9-Z 3/4.9. 9 CONTAINMENT VENTILATION ISOLATION SYSTEM........... . 8 3/4 9-2 3/4.9. lO and 3/4.9.% MATER LEVEL - REACTOR VESSEL AHO H9f S PENT FUEL POQLSe ~ ~ o e ~ ~ o o ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ o ~ o ~ o ~ 8 3/4 9-3 3/4.9.12 FUEL HANOLIHG BUILDING EMERGENCY EXHAUST SYSTEM.... 8 3/4 9-3 SHEARQN HARRIS - UNIT 1 XV
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PLANT SYSTBIS 4.7. 2 AR T P RATURE MONITORING ~ DEW:TED LIMITINQ CONOITION FOR OPERATION
- 3. 7. The temperature of each area shown in Table 3. 7-6 shall not be ceeded for mor than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or by more than 30 F.
APPLICABILI . Whenever the equipment fn an affected area fs quired to be IWEKIB.
ACTION:
With one or moreureas exceeding the temperature limit(s) shown in Table 3.7-6 for more than 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />s~prepare and submit to the Coaiaissfon within 3Q &ys, pursuant'o Specification 6.9.2, a Special Report that prov~es a~cord of the cumulative time and the amount by which the teaqserature in the affected area(s) exceeded the limit(s) andean analysis to demonstrate the continued OPERABILITY of the aff cted equipment. The provisions of Soecifi-cations 3.0.3 and 3..4 are not ap icable.
With one or mor areas exceeding the t rature limit(s) shown in.
Table 3.7-6 more than 30 F, prepare and mit a Special Report as requi by ACTION a. above and wfthfn 4 h rs either restore the a a s) to ~ithin the temperature limit(s) o declare the equip-men n the affected area(s) inoperable.
SURVE CE RE UIREHENTS
.7.12. The temperature in each of the areas shown in Table 3.7-6 shall be determined to be ~ithin its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SHEARON HARRIS " UNIT 1 3/4 7-28
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TABLE 3.7-6 AREA TEMPERATURE 'fONITORINC ' NLETED MAXIMUM TEMPERATURE LIMIT ('F)
R CTOR AUXILIARY BUILDING e
Control Room Envelope (El 305') 85
- 2. Process I&C, Room (El 305') 8
- 3. od Control Cabinets Area (El 305') 4
- 4. A Battery Rooms (EL 286') 85
- 5. A&B uitchgear Rooms (El 286') 90
- 6. Main team, Feedwater Pipe Tunnel (EL 286' 261') 116
- 7. SA&SB E'electrical Penetration Areas (EL 261' 286'), L04 S. Area eieh+MCC 1A35SA aed 1B35SB (S1 26)') 104
- 12. Service Water Booster Pump LB-SB (El 236~') 104
- 13. Mechanical and ELect ical Penetration Areas (El 236') 104
- 14. Containment Spray Add@ ive 'Tank, andrH&V Equipmenc Area (El 216') 104
- 15. Trains A&B Containment Sp yPump', RHR Pump, H&V Equipment Areas (EL 190') 104 FUEL HANDLING BUILDING 16 'rains A&B Emergency Exhaust SystemxAreas (El 261') 104
- 17. DeLeted
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WASTE PROCESSINC BUILDING
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- 18. 'H&V Equipment,Room (El 236') 104 MISCELLANEOUS
- 19. Tank Area (El 236') 104
- 20. Diesel Fuel Oil Storage Building (EL 242') 122
- 21. Emergency Service Mater Electrical Equipmenc Room 104
- 22. Emergency Service Mater Pump Room 22 23 'DELETED 24 LA-SA & 1B-SB H&V Equipmenc Rooms (El '292') 122 LA-SA & IB-SB H&V Equipmenc Rooms (EL 280') 118
- 27. 1A-SA & 1B-SB Diesel. Generator Rooms (El 261') L20 SHEARON HARRI.S " UNIT 1 3/4 7"29 Amendmenc No.
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IHSTRUMEHTATIQN BASES 3/4. 3. 3. 11 RAOIQACTIVE GASEOUS EFFLUEHT MONITORING IHSTRUMEHTATION The radioactive gaseous effluent instrumentation is provided ta monitor and control, as applicable, the releases of radioactive materials in gaseous efflu-ents during actual or potential rel'eases of gaseous effluents. The Alarm/Trip Setpoints far ttiesa instruments shall be calculated and adjusted fn accordance with the methodology and parameters fn the QOQl to ensure that the aIarm/trip will occur prior ta exceeding the limits of 10 CFR Part 20. Thfs instrmenta-tfan also fncludes provisions far monitoring (and controlling) the concentrations of potentially explosive gas mixtures fn the GASEOUS RAGMASTE TREATMEHT SYSTBl.
The OPERABILjTY and use of this instrumentatfon fs consistent with the requB'e-ments of General Oesfgn Criteria 60, 63, and 64 of Appendix A ta 10 CFR Part 50.
The sensitivity. of any noble gas activity monitors used ta shaw compliance with the gaseous effluent release requirements of Specificatfan 3.11.2.2 shall be such that cancentrations as Iaw as 1 x M- pCi/mI are measurable.
3/4.3.4 iilRBZHE OVERSP'E PROTECTION
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This specification is wJ>>
ravi ded Ca.ensure that the tu ine overspeed p taction fnstrumerCation and e turb ine speed control valve a re OPERABLE an will pro-tect the'turbine fram excessive overspeed. Protect:ion fram turbin excessive averspeed fs requir'ed since excessive overspeed pf the turbine calved generate potentially damaging missiles wtvfch cauld fmpaW and damage safe~related ca anerits, equfpment or structures.
SHEARON HARRXS - UNIT 1 & 3/4 3-6
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PLANT SYSTEHS BASES SNUBBERS Continued)
The service life of a snubber is established via manufacturer input and infor mation through consideratfon of the snubber service conditions and associated installation and maintenance records (newly installed snubbers, seal replaced, spring replaced, in high radiation area, in high temperature area, etc. ). The requirement ta monitor the snubber service life is included to ensure that the snubbers periodically undergo a perfonnance evaluation in view of their age and operating conditions. These records wi 11'rovide statistical bases for future cansideratian of snubber service life.
3/4. 7-. 9 SEALED SOURCE CONTAMINATION The sources requiring leak tests are specified in 10 CFR 31.5(c)(Z)(ff). The limitation on remauable contamination is required by 10 CFR 31.5(c)5. This Iimitatian will ensure that leakage from Byproduct, Source, and Special Nuclear Material saurces will not exceed allowable intake values.
Sealed sources are classified inta three groups according to their use, with Surveillance Requirements commensurate with the prababflity of damage ta a source in that group. Those saurces that are frequently handled are required to be tested more often than thase that. are not. Sealed sources that are con-tinuously enclosed within a shielded mechanism (i.e., sealed sources. within.
radiation monitoring or boron measuring devices) are considered to be stared and need not be tested unless they are removed from the shielded mechanism.,
3/4.7.10 DELETED 3/4.7.11 DELETED 3/4.7.12
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AREA TEH temperatures./Exposure RATURE HONIT RING taepera ure iimitatipns ensure t t safety-re'laP'ed equipment ill nnt ected tdtemperatureslin excess np their envirnnmenta1 qual ir cation cause a lose'af its OPERABILITY.
ta jexcessfve temperatures aVfawance $ 6r f nstrument/errors.
/
3.4. 7. 13 The OPERABILITY of I
The+emperature ESSENTIAL SERVICES CHILLED WATER SYSTEH I
may lf its da the Emergency Service Chilled Water System ensures sufficient caolfng capacfty fs available for continued
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an that operation af safety re<ated equipment durfng narmal and accident conditions. The redundant coalfng capacity af this system, assuming a single failure, is consistent with the assumptions used fn the safety analyses.
SHEARON HARRIS - UNIT 1 B 3/4. 7-5
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RAOIOACTIVE EFFLUENTS BASES EXPLOSIVE GAS MIXTURE Continued) prov'ides assurance that the releases of radiaactive materials will be controlled in conformance with the requirements cf General Oesign Criterion 60 of Appen-dix A tc LO.CFR Part 5Q.
3/4.11.2.6 ( S STORAGE TANKS y
'he tan include in this speciPcati on are those /
s for which the uanti of rad'oactivity~antained is nest Iimited directly er'ndirectly by another Tech cal Specification. Restricting the quantity>cf radioactivity contained in each gas storage tank provides assurance that fii the event cf an. uncontrolled re/ease of tNe tank's contents, the resulting w'ncle body exposure.to a MEMBER F THE PUBLIC at the neares't SITE HQUNOARY wi1%'nct exceed 0.5 rem. This is cansis en with Standard review Plan Hi.3, Br/nch Technical Position ETSB 11"5, "Postulated Radioactive releases Oue to a Waste Gas System Leak or Failure," fn NUREG-0800, July 1981.
3/4. 11. 3 SQLIO RAOIOACTIVE WASTES This specification implements the requirements af 10 CFR 50.36a, LO CFR 61, and General Oesign Criterion 60 of Appendix A ta LO CFR Part 50. The process param-eters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited ta, waste type, waste pH, waste/lfquid/SOLIDIFICATION agent/
catalyst ratios, waste oil content, waste principal chemical constituents, and mixing and curirg times.
3/4.11;4 TOTAL OOSE This specification is provided ta meet the dose limitations of LO CFR Part 190 that have been incarparated inta 10 CFR Part 20 by 46 FR 18525. The specifica-tion requires the preparation and submittal cf a Special Report whenever the calculated doses due ta veIeases of radioactivity and to radiation fram uranium fuel cycle sources exceed 25 mrems ta the whale body or any organ, except the thyroid, which shall be limited ta less than cr equal ta 75 mrems. For sites containing up to four reactors, it is highly unlikely that the resultant dose ta a MEMBER OF THE PUBLIC will exceed the dase limits cf 40 CFR Part 190 ff the individual reactors remain within twice the dase design objectives of Appendix I, and ff direct radiation doses,from the units and from outside storage tanks are kept smaII. The Special Report will describe a course cf action that should result fn the Ifmftatfan of the annual dose to a HEHBER OF. THE. PUBLIC ta.within the 40 CFR Part 190 limits. For the .purposes of the Special Report, it may'be assumed that the dose commitment to the MEHBER of the PUBLIC fram other uraniu~
fuel cycle sources is negligible, with the exception that dose contributians fram other. nuclear fuel cycle facilities at the same site or within a radius cf 8 km must be considered. If the dose ta any MEHBER OF THE PUBLIC fs estimated to exceed Che requirements of 40 CFR Part L90, .the Special Report with a request far a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), fn accordance with the provisions of 40 CFR'90.L3. and 10 CFR 20.405c, is considered ta ba a timely request and ful-fills tea requfreaants af 40 CFR Part. 190 until NRC staff action is ccmpleted-SHEARON HARRIS UNIT 1 B 3/4 11-5
'LIMITING CONDITIONS
~ roon FOR OPERATION AND SURVEILLANCE REQUIREMENTS
~
SECTION PAGE TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS.................................... 3/4 3-51 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS.................... 3/4 3-54 Movable Incore Detectors................................ 3/4 3-56 Seismic Instrumentation................................. 3/4 3-57 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION................ 3/4 3-58 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................ 3/4 3-59 Meteorological Instrumentation.....................;.... 3/4 3-60, TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION......... 3/4 3-61 TABLE 4.3-5 METEOROLOGICAL HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................ 3/4 3-62 Remote Shutdown System..... ............................ 3/4 3-63 TABLE 3.3-9 REMOTE SHUTDOWN SYSTEM............................ 3/4 3-64 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................... 3/4 3-65 Accident Monitoring Instrumentation..................... 3/4 3-66 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION............. 3/4 3-68 TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE
.REQUIREMENTS.....;...................................... 3/4 3-70 TABLE 3.3-11 (DELETED)................................... 3/4 3-73 Hetal Impact Monitoring System.......................... 3/4 3-74 Radioactive Liquid Effluent Monitoring Instrumentation.. 3/4 3-75 TABLE 3.3-12 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUHENTATION...................-.................. 3/4 3-76 TABLE 4.3-8 RADIOACTIVE LIOUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............... 3/4 3-79 Radioactive Gaseous Effluent Monitoring Instrumentation. 3/4 3-82 TABLE 3.3-13 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION................... 3/4 3-83 TABLE 4.3-9 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUHENTATION SURVEILLANCE REQUIREMENTS............... 3/4 3-86 3/4.3. 4 ( D E L ET E D ) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 3-89 I SHEARON HARRIS - UNIT 1 vi Amendment No.
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION..... 3/4 7-10 3/4.7.3 COMPONENT COOLING WATER SYSTEM...................... 3/4 7-11 3/4.7.4 EMERGENCY SERVICE WATER SYSTEM...................... 3/4 7-12 3/4.7.5 ULTIMATE HEAT SINK..........................-..-.... 3/4 7-13 3/4.7.6 CONTROL ROOM EMERGENCY FILTRATION SYSTEM............ 3/4 7-14 3/4.7.7 REACTOR AUXILIARY BUILDING (RAB) EMERGENCY EXHAUST SYSTEM.......................... 3/4 7-17 3/4.7.8 SNUBBERS............................................ 3/4 7-19 FIGURE 4.7-1 (DELETED)....................................... 3/4 7-24 3/4.7.9 SEALED SOURCE CONTAMINATION......................... 3/4 7-25 3/4.7.10 ( DELETED) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-27 TABLE 3.7-3 ( DELETED) ~ ~ ~ ~ ~ ~ ~ ~ ~ o o ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-27 TABLE 3.7-4 (DELETED)...............................-............ 3/4 7-27 TABLE 3.7-5 (DELETED)............................................ 3/4 7-27 3/4.7.11 ( DELETED)o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 7-27 3/4.7.12 ( DELETED)........................................... 3/4 7-28 TABLE 3.7-6 ( DELETED)t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
o'
~ ~ ~ ~ ~ ~ ~ 3/4 7-29 3/4.7.13 ESSENTIAL SERVICES CHILLED WATER SYSTEM............. 3/4 7-30 3 4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 0 perating............................................... 3/4 8-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE................... 3/4 8-10 S hutdown................................................ 3/4 8-11 3/4.8.2 D.C. SOURCES 0 peratsng.... 3/4 8-12 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS..........-..... 3/4 8-14 S hutdown................................................ 3/4 8-15 3/4.8.3 ONSITE POWER DISTRIBUTION 0 perating...............................................
4 3/4 8-16 S hutdown.................. 3/4 8-18 SHEARON HARRIS - UNIT 1 Amendment No.
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3 4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS C oncentratson...........................................
J ~
3/4 11-1 TABLE 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS P ROGRAH ~ ~ o ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 11-2 D ose.................................................... 3/4 11-5 Liquid Radwaste Treatment System........................ 3/4 11-6 Liquid Holdup Tanks..................................... 3/4 11-7 3/4.11.2 GASEOUS EFFLUENTS 0 ose Rate............................................... 3/4 11-8 TABLE 4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS P ROGRAMt ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 11-9 D ose - Noble Gases...................................... 3/4 11-12 Dose - Iodine-131, Iodine-133, Tritium, and Radioactive Material in Particulate Form............................ 3/4 11-13 Gaseous Radwaste Treatment System....................... 3/4 11-14 E xplossve n Gas Mixture...................................
aa ~
3/4 11-15
( DELETED)o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~
o' ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 11"16 I 3/4.11.3 SOLID RADIOACTIVE WASTES................................ 3/4 11-17 3/4.11.4 TOTAL DOSE.............................................. 3/4 11-19 3 4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PROGRAM...................................... 3/4 12-1 TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAH....... 3/4 12-3 TABLE 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES................................ 3/4 12-9 TABLE 4.12-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 3/4 12-10 3/4.12.2 LAND USE CENSUS......................................... 3/4 12-13 3/4. 12.3 INTERLABORATORY COMPARISON PROGRAM...................... 3/4 12-14 SHEARON HARRIS - UNIT 1 xii Amendment No.
INDEX 3.0/4.0 BASES SECTION PAGE 3 4.0 APPLICABILITY.............................................. B 3/4 0-1 3 4.1 REACTIVITY CONTROL SYSTEMS 3/4.1. 1 BORATION CONTROL......................................... B 3/4 1-1 3/4.1.2 BORATION SYSTEMS......................................... B 3/4 1-2 3/4. 1.3 MOVABLE CONTROL ASSEMBLIES............................... B 3/4 1-3 3 4. 2 POWER DISTRIBUTION LIMITS.................................. 8 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE............................;....... B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, AND RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR....... B 3/4 2-2 FIGURE 8 3/4.2-1 (DELETED)....................................... B 3/4 2-3 3/4.2.4 QUADRANT POWER TILT RATIO................................ B 3/4 2-5 3/4.2.5 DNB PARAMETERS.......-................................... B 3/4 2-6 3 4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION...... B 3/4'3-1 3/4.3.3 MONITORING INSTRUMENTATION............................... B 3/4 3-3 3/4.3.4 (DELETED).................................. ~ ... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 3-6 I 3 4.4 REACTOR COOLANT SYSTEM 3/4.4. 1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION............ B 3/4 4-1 3/4.4.2 SAFETY VALVES............................................ B 3/4 4-1 3/4.4.3 PRESSURIZER.............................................. B 3/4 4-2 3/4.4.4 REL I EF VALVES~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ 8 3/4 4-2 3/4.4.5 STEAM GENERATORS......................................... ~
B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE........................... B 3/4 4-3 3/4.4.7 C HEMISTRY.................................
h B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 4-5 3/4.4.9 PRESSURE/TEMPERATURE LIMITS........................... - .. B 3/4 4-6 SHEARON HARRIS - UNIT 1 X111 Amendment No.
INDEX BASES SECTION PAGE 3 4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................................. B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION.......... B 3/4 7-2 3/4.7.3 COMPONENT COOLING WATER SYSTEM........................... B 3/4 7-3 3/4.7.4 EMERGENCY SERVICE WATER SYSTEM........................... B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK....................................... B 3/4 7-3 3/4.7.6 CONTROL ROOM EMERGENCY FILTRATION SYSTEM................. B 3/4 7-3 3/4.7.7 REACTOR AUXILIARY BUILDING EMERGENCY EXHAUST SYSTEM...... B 3/4 7-3 3/4.7.8 S NUBBERS ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 7-4 3/4.7.9 SEALED SOURCE CONTAMINATION.............................. B 3/4 7-5 3/4.7.10 ( DELETED)................................................ B 3/4 7-6 3/4.7.11 ( DELETED) ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 7-6 3/4.7.12 ( DELETED) ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 7-6 )
3/4 7.13 ESSENTIAL SERVICES CHILLED WATER SYSTEM.................. B 3/4 7-6 3 4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, AND 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION............................... B 3/4 8-1 3/4.8.4 ELECTRICAL E(UIPMENT PROTECTIVE DEVICES.................. B 3/4 8-3 3 4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION...................................... B 3/4 9-1 3/4.9.2 INSTRUMENTATION.......................................... B 3/4 9-1 3/4.9.3 D ECAY TIME ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ i ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ B 3/4 9-1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS........................ B 3/4 9-1 3/4.9.5 COMMUNICATIONS...................;....................... B 3/4 9-1 3/4.9.6 REFUELING MACHINE........................................ B 3/4 9-2 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING.................... B 3/4 9-2 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION............ B 3/4 9-2 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM................. B 3/4 9-2 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND NEW AND SPENI FUEL POOLS......................................... B 3/4 9-3 3/4.9.12 FUEL HANDLING BUILDING EMERGENCY EXHAUST SYSTEM......... B 3/4 9-3 SHEARON HARRIS - UNIT 1 XV Amendment No.
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INSTRUMENTATION 1
3 4.3.4 TURBINE OVERSPEED PROTECTION - DELETED SHEARON HARRIS - UNIT I 3/4 3-89 Amendment No.
PLANT SYSTEMS 3 4.7. 12 AREA TEMPERATURE MONITORING - DELETED SHEARON HARRIS - UNIT 1 3/4 7-28 Amendment No.
TABLE 3.7-6
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p AREA TEMPERATURE MONITORING - DELETED SHEARON HARRIS - UNIT I 3/4 7-29 Amendment No.
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RADIOACTIVE EFFLUEN GAS STORAGE TANKS - DELETED SHEARON,HARRIS - UNIT 1 3/4 ll-l6 Amendment No.
INSTRUMENTATION V
BASES 3 4.3.3.11 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The Alarm/Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the GASEOUS RADWASTE TREATMENT SYSTEM. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any noble gas activity monitors used to show compliance with the gaseous effluent release requirements of Specification 3. 11.2.2 shall be such that concentrations as low as 1 x 10'Ci/ml are measurable.
3 4.3.4 DELETED SHEARON HARRIS - UNIT 1 8 3/4 3-6 Amendment No.
PLANT SYSTEMS BASES SNUBBERS Continued The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubbers, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide
~ statistical bases for future consideration of snubber service life.
3 4.7.9 SEALED SOURCE CONTAMINATION The sources requiring leak tests are specified in 10 CFR 31.5(c)(2)(ii). The 1 imitation on removable contamination is r equired by 10 CFR 31.5(c) 5. This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values.
Sealed sources are classified into three groups according to their use, with Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources that are frequently handled are required to be tested more often than those that are not. Sealed sources that are con-tinuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.
3 4.7.10 DELETED 3 4.7.11 DELETED 3 4.7. 12 DELETED 3.4.7.13 ESSENTIAL SERVICES CHILLED WATER SYSTEM The OPERABILITY of the Emergency Service Chilled Water System ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the safety analyses.
SHEARON HARRIS - UNIT 1 B 3/4 7-5 Amendment No.
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RADIOACTIVE EFFLUEN BASES EXPLOSIVE GAS MIXTURE Continued provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.
3 4.11.2.6 DELETED 3 4.11.3 SOLID RADIOACTIVE WASTES This specification implements the requirements of 10 CFR 50.36a, 10 CFR 61, and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to, waste type, waste pH, waste/liquid/SOLIDIFICATION agent/catalyst ratios, waste oil content, waste principal chemical constituents, and mixing and curing times.
3 4. 11.4 TOTAL DOSE specification is provided to meet the dose limitations of 10 CFR Part 190
'his that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the prepar ation and submittal of a Special Report when-ever the calculated doses due to releases of radioactivity and to radiation from uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units and from outside storage tanks are kept small. The Special Report will describe a course of action th'at should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR Part 190 limits. For the purposes of the Special Report, it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear fuel cycle facilities at the same site or within a radius of 8 km must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR Part 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR Part 190 have not already been corrected), in accordance with the provisions of 40 CFR 190. 11 and 10 CFR 20.405c, is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 until NRC staff action is completed.
SHEARON HARRIS - UNIT 1 B 3/4 11-5 Amendment No.
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