ML20035D252
| ML20035D252 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/05/1993 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | GPU Nuclear Corp, Jersey Central Power & Light Co, Metropolitan Edison Co, Pennsylvania Electric Co |
| Shared Package | |
| ML20035D253 | List: |
| References | |
| GL-89-001, DPR-50-A-173 NUDOCS 9304120348 | |
| Download: ML20035D252 (30) | |
Text
. - _ _ _ _
/pa arctq'o UNITED STATES
[" 3 m "g NUCLEAR REGULATORY COMMISSION
, a WASHINGTON D. C,20555 1., -...../
METROPOLITAN EDISON COMPANY r
JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR CORPORATION DOCKET NO. 50-289 I
THREE MILE ISLAND NUCLEAR STATION. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.173 License No. DPR-50 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by GPU Nuclear Corporation, et al.,
(the licensee) dated May 19, 1992, as supplemented by letters dated November 30, 1992, January 29, and February 12, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements
)
have been satisfied, i
9304120348 930405 ADOCKOSOOg9 PDR P
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2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.c.(2) of Facility Operating License No.
DPR-50 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.173, are hereby incorporated in the license.
GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance, to be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULAT RY COMMISSION l
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/doh
. Stolz, Director Pr ject Directorate I-4 p vision of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 5,1993
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ATTACHMENT TO LICENSE AMENDMENT NO. 173 FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Replace the following pages of the Appendix A Technical Specifications with the attached page. The revised pages are identified by an amendment number and contains vertical lines indicating the area of change.
Remove Insert i
i v
v vi vi 1-6 1-6 1-7 1-7 1-8 1-8 1-9 3-96 3-96 3-97 3-97 3-98 3-98 3-100 3-100 3-106 3-106 3-111 3-111 3-118 3-118 3-120 3-120 3
3-121 i
3-122 3-122 i
3-123 3-124 3-125 3-125 3-126 3-126 3-127 3-127 4-99 4-99 4-105 4-105 4-107 4-107 4-108 4-108 4-117 4-117 4-118 4-118 4-119 4-120 4-121 4-121 4-122 4-122 6-11 6-11 6-lla 6-17 6-17 6-18 6-18 6-19 6-19 6-19a 6-20 6-20 6-21 6-21 6-22 6-22 6-23 6-23 6-24 6-24 6-25 i
TABLE OF CONTENTS Section Paoe TECHNICAL SPECIFICATIONS 1
DEFINITIONS 1-1 1.1 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 ptartup 1-2 1.2.10 Avg 1-2 1.2.11 Heatup-Cooldown Mode 1-2 1.2.12 Station, Unit, Plant, and Facility 1-2 1.3 OPERABLE l-2 1.4 PROTECTIVE 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 Trip Test 1-3 q
1.5.2 Channel Test 1-3 l.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 Power Tilt 1-5 1.6.2 Axial Power Imbalance 1-5 1.7 CONTAINMENT INTEGRITY 1-5 1.8 FIRE SUPPRESSION WATER SYSTEM 1-5 1.9 DELETED l-6 1.10 DELETED l-6 1.11 DELETED l-6 1.12 DOSE EQUIVALENT I-131 1-6 1.13 SOURCE CHECK 1-6 1.14 SOLIDIFICATION - DELETED 1-6 l
1.15 0FFSITE DOSE CALCULATION MANUAL l-6 1.16 PROCESS CONTROL PROGRAM 1-6 1.17 GASEOUS RADWASTE TREATMENT SYSTEM l-6 1.18 VENTILATION EXHAUST TREATMENT SYSTEM 1-7 j
1.19 PURGE-PURGING l-7 i
1.20 VENTING l-7 l
1.21 REPORTABLE EVENT 1-7 1.22 MEMBER (S) 0F THE PUBLIC 1-7 1.23 SUBSTANTIVE CHANGES 1-7 1.24 CORE OPERATING LIMITS REPORT l-8 l
1.25 FRE0VENCY NOTATION 1-8 i
Amendment No. 11, 72, 129, 137, 142, 150, 155,173
-9205270039-920519 PDR ADOCK 05000289 P
1 TABLE OF CONTENTS Section EiLqt 5
DESIGN FEATURES 5-1 5.1 111E 5-1 5.2 CONTAINMENT 5-2 5.2.1 REACTOR BUILDING 5-2 5.2.2 REACTOR BUILDING ISOLATION SYSTEM 5-3 5.3 REACTOR 5-4 5.3.1 REACTOR CORE 5-4 5.3.2 REACTOR COOLANT SYSTEM 5-4 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 5-6 5.4.1 NEW FUEL STORAGE 5-6 5.4.2 SPENT FUEL STORAGE 5-6 5.5 AIR INTAKE TUNNEL FIRE PROTECTION SYSTEMS 5-8 6
ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-1 6.2.1 CORPORATE 6-1 6.2.2 UNIT STAFF 6-1 6.3 UNIT STAFF OVALIFICATIONS 6-3 6.4 TRAINING 6-3 6.5 REVIEW AND AUDIT 6-3 6.5.1 TECHNICAL REVIEW AND CONTROL 6-4 6.5.2 INDEPENDENT SAFETY REVIEW 6-5 6.5.3 AUDITS 6-7 6.5.4 INDEPENDENT ONSITE SAFETY REVIEW GROUP 6-8 6.6 REPORTABLE EVENT ACTION 6-10 6.7 SAFETY LIMIT VIOLATION 6-10 6.8 PROCEDURES AND PROGRAMS 6-11 l
6.9 REPORTING RE0VIREMENTS 6-12 6.9.1 ROUTINE REPORTS 6-12 6.9.2 DELETED 6-14 6.9.3 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6-17 6.9.4 ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 6-18 j
J 6.9.5 CORE OPERATING LIMITS REPORT 6-19 J
6.10 RECORD RETENTION 6-20 6.11 RADIATION PROTECTION PROGRAM 6-22 6.12 HIGH RADIATION AREA 6-22 6.13 PROCESS CONTROL PROGRAM 6-23 6.14 0FFSITE DOSE CALCULATION MANUAL (0DCM) 6-24 j
6.15 DELETED 6-24 6.16 EQST ACCIDENT SAMPLING PROGRAMS 6-24 NUREG 0737 (ll.B.3. II.F.1.2) 6.17 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS 6-25 l
-v-i Amendment No. JJ, (7, 77, 77, JEP, 150,173 4
LIST OF TABLES TABLE TITLE
_ pag 1 A
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 3-15a Primary Coolant System and LPIS 3.5-1 Instruments Operating Conditions 3-29 3.5-1A DELETED 3.5-2 Accident Monitoring Instruments 3-40c 3.5-3 Post Accident Monitoring Instrumentation 3-40d 3.21-1 Radioactive Liquid Effluent Monitoring 3-97 Instrumentation 3.21-2 Radioactive Gaseous Process and Effluent 3-101 Monitoring Instrumentation 3.23-1 DELETED 3.23-2 DELETED 4.1-1 Instrument Surveillance 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 1
4.21-1 Radioactive Liquid Effluent Monitoring 4-88 Instrumentation Surveillance Requirements 4.21-2 Radioactive Gaseous Effluent Monitoring 4-91 Instrumentation Surveillance Requirements 1
4.22-1 Radioactive Liquid Waste Sampling &
4-96 Analysis Program
{
4.22-2 Radioactive Gaseous Waste Sampling &
4-102 Analysis Program i
4.23-1 DELETED vi Amendment No. 59, 72, 100, 106, 118, 137, 142, 147, 150,173 j
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1.9 DELETED 1.10 DELETED l
1.11 DELETED 1.12 DOSE EQUIVALENT I-131 The DOSE EQUIVALENT I-131 shall be that concentration of I-131 i
(microcurie / gram) which alone would produce the same thyroid dose as the i
quantity and isotopic mixture of I-131, I-132, I-133, 1-134 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 l
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 DELETED 1.15 0FFSITE DOSE CALCULATION MANUAL (ODCM)
The 0FFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluent, in the calculation of gaseous and liquid effluent I
monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radiological Environmental Monitoring Program required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual j
Radiological Environmental Operating and Annual Radioactive Effluent Release j
Reports required by Specifications 6.9.3 and 6.9.4.
1.16 PROCESS CONTROL PROGRAM (PCP)
The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.
1.17 GASEOUS RADWASTE TREATMENT The GASEOUS RADWASTE TREATMENT SYSTEM is the system designed and installed to reduce radioactive gaseous effluent 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.
1-6 Amendment No. 72, 137,173
1.18 VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluent by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the purpose of removing iodine 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.
1.19 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.
1.21 REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in 10 CFR 50.73.
1.22 MEMBER (S) 0F THE PUBLIC MEMBER (5) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the GPU System, GPU contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries.
1.23 SUBSTANTIVE CHANGES SUBSTANTIVE CHANGES are those which affect the activities associa'ted with a document or the document's meaning or intent.
Examples of non-substantive changes are: (1) correcting spelling; (2) adding (but not deleting) sign-off spaces; (3) blocking in notes, cautions, etc.; (4) changes in corporate and personnel titles which do not reassign responsibilities and which are not referenced in the Appendix A Technical Specifications; and (5) changes in nomenclature or editorial changes 'which clearly do not change function, meaning or intent.
1-7 Amendment No. 72, 137, 141, 150, 155, 157, 158,173
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1.24 CORE OPERATING LIMITS REPORT-The CORE OPERATING LIMITS REPORT is a TMI-1 specific document that provides core i
operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.5.
Plant operation within these operating limits is i
addressed in individual specifications.
1.25 FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements.
shall correspond to the intervals defined in Table 1.2.
All Surveillance Requirements shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval.
TABLE 1.2 FRE0VENCY NOTATION NOTATION FRE0VENCY S
Shiftly (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 />)
I 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 i
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 18 months l
1-8 Amendment No. 72, 137, 155,173 ne
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1.25 FREQUENCY NOTATION Bases Section 1.25 establishes the limit for which the specified time interval for Surveillance Requirements may be extended, it permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g.,
transient conditions or other ongoing surveillance or maintenance activities.
It also provides flexibility to accommodate the length of a fuel cycle for surveillance that are performed at each refueling outage and are specified with a fuel cycle length surveillance interval.
It is not intended that this provision be used repeatedly as a convenience to extend surveillance interval s beyond that specified for surveillance that are not performed during refueling outages. The limitation of Section 1.25 is based on engineering judgement and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.
1 1-9 Amendment No. 72, 137, 155,173
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3.21 RADI0 ACTIVE EFFLUENT INSTRUMENTATION 3.21.1 RADI0 ACTIVE LIOUID EFFLUENT INSTRUMENTATION LIMITING CONDITION FOR OPERATION i
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 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 (ODCM).
APPLICABILITY:
At all times
- ACTION:
a.
With a radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluent monitored by the affected channel or declare the channel inoperable.
b.
With less than the minimum number of radioactive liquid effluent
+
monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.21-1.
Exert best efforts to return the instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Annual Effluent Release Report why the inoperability was l
not corrected in a timely manner.
- 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.
l
- 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 c
The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluent 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.
i 3-96 knendment No. 72, ES,137,173
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TABLE 3.21-1 MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION 1.
Gross Radioactivity Monitors Providing Automatic Termination of Release a.
Unit 1 Liquid Radwaste Effluent 1
18 Line (RM-L6) b.
IWTS/IWFS discharge line (RM-L12) 1 20 2.
[ DELETED BY AMENDMENT NO. 88]
3.
Flow Rate Measurement Devices a.
Unit 1 Liquid Radwaste 1
21 Effluent Line (FT-84) b.
Station Effluant Discharge 1
21 (FT-146) 1 3-97 Amendment No. 72, 88, 173
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1 3-98 Amendment No. 72, 88,173
3.21.2 RADI0 ACTIVE GASE0US PROCESS AND EFFLUENT MONITORING INS 1?yMENTATION LIMITING CONDITION FOR OPERATION 3.21.2 The radioactive gaseous process and effluent monitoring instrumentation channels shown in Table 3.21-2 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.22.2.1 are not exceeded.
The alarm / trip setpoints of these channels shall be determined in accordance with the OFFSITE DOSE CALCULATION KANUAL (0DCM).
APPLICABILITY:
As shown in Table 3.21-2.
ACTION:
)
i a.
With a radioactive gaseous process or effluent monitoring i
instrumentation channel alarm / trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive effluent monitored by the affected channel or declare the channel inoperable.
b.
With less than the minimum number of radioactive gaseous process or effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.21-2.
Exert best efforts to return the instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Annual Effluent Release Report l
why the inoperability was not corrected in a timely manner.
BASES The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluent during actual or potential releases.
The ala'rm/ 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.
The low range condenser offgas noble gas activity monitors also provide data for determination of steam generator primary to secondary le?.kage rate.
Channel operability requirements are based on an ASLB Order No. LBP-84-47 dated October 31, 1984, and as cited in 20 NRC 1405 (1984).
i 3-100 Amendment No. 77, 19), 177, 157,173
3.22 RADI0 ACTIVE EFFLUENT 3.22.1 LIOUID EFFLUENT
)
3.22.1.1 CONCENTRATION LIMITING CONDITION FOR OPERATION 3.22.1.1 The concentration of radioactive material released at anytime fro the unit to unrestricted areas (see Figure 5-3) shall be limited to the radionuclides other than dissolved or entrained noble gase m
entrained noble gases, the concentration shall be limited to 3 x 10'p1ved o or For diss total activity.
uti/cc APPLICABILITY: At all times ACTION:
With the concentration of radioactive material release within the above limits. unrestricted areas exceeding the above limits, immed ration BASES This specification is provided to ensure that the concentration of materials released in liquid waste effluent from the unit to unrestrict d radioactive will be less than the concentration levels specified in 10 CFR Part 2 Appendix B, Table II.
e areas levels of radioactive materials in bodies of water outside the sitT result in exposures with (1) the Section ll. A design objectives of App e will not 10 CFR Part 50, to a MEMBER OF THE PUBLIC and (2) the limits of 10 endix I, i
20.106 (e) to the population.
upon the assumption the Xe-135 is the controlling radioisotope and
{
(submersion) was converted to an equivalent concentration in w t n air l
methods described in International Commission on Radiological Prot a er using the Publication 2.
j ection(ICRP)
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3-106 Amendment No. 72, 137, 149,173
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RADIOACTIVE EFFLUENT 3.22,2 GASE0US EFFLUENT 3.22.2.1 DOSE RATE LIMITING CONDITION FOR OPERATIONS 3.22.2.1 The dose rate due to radioactive materials released in gaseous effluent 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:
With the release rate (s) exceeding the above limits, imediately decrease the release rate to comply with the above limit (s).
BASES The specification is provided to ensure that the release rate at anytime at the site boundary from gaseous effluent 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 provide reasonable assurance that radioactive material discharged in gaseous effluent will not result in the exposure of a MEMBER OF THE PUBLIC in an unrestricted area, either within or outside the site boundary, to annual average concentrations exceeding the limits specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)).
For MEMBERS OF THE PUBLIC who may at times be within the site boundary, the occupancy of the MEMBER OF THE PUBLIC 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 a MEMBER OF THE PUBLIC 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 a child via the cow-milk pathway to less than or equal to 1500 mrem / year for the nearest cow to the pl ant.
3-111 Amendment No. 72, 137.173
3.22.3 SOLID RADI0 ACTIVE WASTE LIMITING CONDITION FOR OPERATION i
DELETED I
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Amendment No. 72, 129,173
i 3.23 RADIOLOGICAL ENVIRONMENTAL MONITORING 3.23.1 MONITORING PROGRAM
[
f LIMITING CONDITION FOR OPERATION I
DELETED t
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1 3-120 (3-121 deleted)
Amendment No. 72,173
TABLE 3.23-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM DELETED j
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i 3-122 (3-123 thru 3-124 deleted)
Amendment No. 72,173
RADIOLOGICAL ENVIRONMENTAL MONITORING 3.23.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION i
DELETED l
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3-125 Amendment No. 72,173
TABLE 3.23-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES i
DELETED t
L e
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P h
3-126 i
Amendment No. 72,173
RADIOLOGICAL ENVIRONMENTAL MONITORING 3.23.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION DELETED r
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3-127 Amendment No. 72,173 1
.. =..
c.
To be representative of the' quantities and concentrations of radioactive-
-l materials. in liquid effluent, samples shall be collected continuously in t
proportion to the rate of flow of the effluent stream.
Prior to analyses, all samples.taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.
i 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 be thoroughly mixed, by a method described in the ODCM, to assure l
representative sampling.
'l e.
A continuous release is the discharge of liquid wastes of a non-discrete i
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-l exclusively are the following radionuclides: Mn-54, Fe-59, Co-58,-
C0-60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144.
This list does not mean that only these nuclides are to be considered. Other gamma l
peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent l
l Release Report pursuant to TS 6.9.4.
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Amendment No. 72, 137,173 j
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Table 4.22-2 (Continued)
J 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 after removal from sampler).
I 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 tim 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 gamma 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, C0-58, C0-60, Zn-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 considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radioactive Effluent l
Release Report pursuant to TS 6.9.4.
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.
i i.
Gross Alpha, Sr-89, and Sr-90 analyses do not apply to the Fuel Handling Building ESF Air Treatment System.
j.
If the Condenser Vent Stack Continuous Iodine Sampler is unavailable, then alternate sampling equipment will be placed in service within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or a report will be prepared and submitted within 30 days from the time the sampler is found or made inoperable which identifies (a) the cause of the inoperability, (b) the action taken to restore representative sampling capability, (c) the action taken to prevent recurrence, and (d) quantification of the release via the pathway during the period and comparison to the limits prescribed by TS 3.22.2.1.b.
k.
Applicable only when condenser vacuum is established.
4-105 Amendment No. 72, 122, 130, 137, 161,1'73
i 4.22.3 SOLID RADIOACTIVE WASTE SURVEILLANCE RE0VIREMENTS 4.22.3.1 SOLID RADWASTE SYSTEM DELETED 4.22.3.2 PROCESS CONTROL PROGRAM DELETED I
i 4-107 Amendment No. 72, 137,173
4.22.4 TOTAL DOSE SURVEILLANCE REOUIREMENT 4.22.4.1 DOSE CALCULATION Cumulative annual 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, including direct radiation contributions from the Unit and from outside storage tanks, and in accordance with the methodology and parameters contained in the ODCM.
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Amendment No. 72, 137,173 1
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e SURVEILLANCE REQUIREMENTS 4.23.1 C:LETED I
4-117 Amendment No. 72,173
TABLE 4.23-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)
DELETED 5
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I 4-118 (4-119 thru 4-120 deleted)
Amendment No. 72,173 i
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SURVEILLANCE REQUIREMENTS i
4.23.2 i
DELETED I
f 4-121 Amendment No. 72,173
SURVEILLANCE REQUIREMENTS 4.23.3 DELETED 1
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4-122 Amendment No. 72,173 l
6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the items referenced below:
a.
The applicable procedures recommended in Appendix
- A" of Regulatory Guide 1.33, Revision 2, February 1978, b.
Surveillance and test activities of equipment that affects 1
nuclear safety and radioactive waste management equipment.
c.
Refueling Operations.
d.
Security Plan Implementation.
e.
Fire Protection Program Implementation.
f.
Emergency Plan Implementation.
g.
Process Control Program Implementation.
j h.
Offsite Dose Calculation Manual Implementation.
i.
Quality Assurance Program for effluent and environmental monitoring using the guidance in Regulatory Guide 4.15, Revision 1.
j.
Plant Staff Overtime, to limit the amount worked by staff performing safety-related functions in accordance with NRC Policy Statement on working hours (Generic Letter No. 82-12).
i 6.8.2 Further, each procedure required by 6.8.1 above, and substantive changes thereto, shall be reviewed and approved as described in 6.5.1 prior to implementation and shall be reviewed periodically as set forth in administrative procedures.
6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided:
a.
The intent of the original procedure is not altered; b.
The change is approved by two members of GPUNC Management Staff qualified in accordance with 6.5.1.14 and knowledgeable in the area affected by the procedure. For changes which may affect the operational status of unit systems or equipment, at least one of these individuals shall be a member of unit management or supervision holding a Senior Reactor Operator's License on the unit.
c.
The change is documented, reviewed and approved as described in 6.5.1 within 14 days of implementation.
6-11 Amendment No. 11, 32, 72, 77, 84, 129, 141, 157,173
6.8.4 Radioloaical Environmental Monitorino Procram A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways.
The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:
(1) Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM, (2) A Land Use Census to ensure that changes in the use of areas at and beyond the SITE B0UNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and (3)
Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.
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Amendment No. 173 6-lla
6.9.3 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6.9.3.1 The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be j
submitted prior to May 1 of each year, j
The Report shall include summaries, interpretations, and an analysis of trends of the results of the Radiological Environmental 1
Monitoring Program for the reporting period. The material provided i
shall be consistent with the objectives outlined in: (1) the ODCM; and, (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.
1 Epig: A single submittal may be made for the station.
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6-17 Amendment No. 59, 64, 72, 77, 108, 117, 129,173 J
~
6.9.4 ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 6.9.4.1 The Annual Radioactive Effluent Release Report covering the operations of the unit during the previous 12 months of operation shall be submitted within 60 days after January 1 of each year.
The Report shall include a summary of the quantities of radioactive liquid anc gaseous effluent and solid waste released from the unit. The material provided shall be: (1) consistent with the objectives outlined in the ODCM and PCP; and, (2) in conformance with 10 CFR 50.36(a) and Section IV.B.1 of Appendix I to 10 CFR Part 50.
Note:
A single submittal may be made for the station. The submittal should combine those sections that are common to both units at the station.
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6-18 Amendment No. 72, 77, J29,137,173
6.9.5 CORE OPERATING LIMITS REPORT 6.9.5.1 The core operating limits addressed by the individual Technical Specifications shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle or prior to any remaining part of a reload cycle.
6.9.5.2 The analyti. cal methods used to determine the core operating limits addressed by the individual Technical Specifications shall be those previously reviewed and approved by the NRC for use at TM1-1, specifically:
(1) BAW-10122A Rev. 1, " Normal Operating Controls," May 1984.
(2)
BAW-10ll6-A, " Assembly Calculations and Fitted Nuclear Data,"
May 1977.
(3)
BAW-10ll7P-A, " Babcock & Wilcox Version of PDQ User's Manual,"
January 1977.
(4) BAW-10ll8A, " Core Calculational Techniques and Procedures,"
December 1979.
(5)
BAW-10124A, " FLAME 3 - A Three-Dimensional Nodal Code for Calculating Core Reactivity and Power Distributions," August 1976.
(6) BAW-10125A, " Verification of Three-Dimensional FLAME Code,"
August 1976.
(7)
BAW-10152A, " NOODLE - A Multi-Dimensional Two-Group Reactor Simulator," June 1985.
(8) BAW-10119, " Power Peaking Nuclear Reliability Factors," June 1977.
(9) BAW-10103, Rev. 3, "ECCS Analysis of B&W's 177-FA Lowered Loop NSS,"
July 1977.
(10) BAW-1915P, " Bounding Analytical Assessment of NUREG-0630 Models on LOCA kw/ft Limits With Use of FLECSET," May 1986.
(11 BAW-10104P-A, Rev. 5,"B&W ECCS Evaluation Model," November 1988.
(12) BAW-10162P-A, " TACO-3 Fuel Pin Thermal Analysis Computer Code,"
November 1989.
6.9.5.3 The core operating limits shall be determined so that all applicable limits (e.g. fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient / accident analysis limits) of the safety analysis are met.
6.9.5.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.
l 6-19 Amendment No. 72,77,129,137,141,149,150,168,173
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6.10 RECORD RETENTION 6.10.1 The following records shall be retained for,at least five years:
1 a.
Records of normal station operation including power levels and '
periods of operation at each power level.
i b.
Records of principal maintenance activities, including inspection, repairs, substitution, or replacement of principal items.of equipment related to nuclear safety.
i c.
All REPORTABLE EVENTS.
d.
Records of periodic checks, tests and calibrations.
i e.
Records of reactor physics tests and other special tests related to nuclear safety.
l f.
Changes to procedures required by Specification 6.8.1.
g.
Records of solid radioactive shipments.
i h.
Test results, in units of microcuries, for leak tests performed on licensed sealed sources.
i.
Results of annual physical inventory verifying accountability of licensed sources on record.
J.
Control Room Log Book.
k.
Shift Foreman Log Book.
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6-20 Amendment No. 77,77,129,JJ7,JJJ,JJp, 150,173
6.10.2 The.folloaing records shall be retained for the duration of Operating License DPR-50 unless otherwise specified in 6.10.1 above, a.
Records and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.
f b.
Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
c.
Routine unit radiation surveys and monitoring records.
d.
Records of radiation exposure history and radiation' exposure i
status of personnel, including all contractors and unit visitors who enter radioactive material areas.
i e.
Records of radio &ctive liquid and gaseous wastes released to the environment, and records of environmental monitoring
- surveys, f.
Records of transient or operational cycles for those facility components which affect nuclear safety for a limited number of transients or cycles as defined in the Final Safety Analysis l
Report.
g.
Records of training and qualification for current members of the unit staff.
h.
Records of in-service inspections performed pursuant to these f
Technical Specifications.
1.
Records of Quality Assurance activities required by the Operational Quality Assurance Plan.
l J.
Records of reviews performed for changes made to procedures or l
equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
l Records of reviews by the Independent Onsite Safety Review k.
Group.
1.
Records of analyses required by the radiological environmental i
monitoring program.
m.
Records of the service lives of all safety related hydraulic snubbers including the date at which the service life commences l
and associated installation and maintenance records.
6.10.3 The following records shall be retained for the duration of the unit i
Operating License:
o.
Records of reviews performed for changes made to the OFFSITE i
DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM.
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6-21 j
Amendment No. 72, 77, J27, J77, JJJ, JJP, Nd. 168,173 f
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1 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation _ exposure.
6.12 HIGH RADI ATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203 (c)(2) of 10 CFR 20:
- a. Each High Radiation Area as defined by paragraph 20.202 (b)(3) shall be barricaded and conspicuously posted as a High Radiation Area, and personnel desiring entrance shall obtain a Radiation Work Permit (RWP).
Any individual or group of individuals entering a High Radiation Area shall (a) use a continuously indicating dose rate monitoring device or (b) use a radiation dose rate integrating device which alarms at a pre-set dose level (entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them), or (c) assure that a radiological control technician provides positive control over activities within the area and periodic radiation surveillance with a dose rate monitoring instrument.
- b. Any area accessible to personnel where a major portion of the body could receive in any one hour a dose in excess of one thousand mrem shall be locked or guarded to prevent unauthorized entry. The keys to these locked barricades shall be maintained under the administrative control of the respective Radiological Controls Supervisor.
The Radiation Work Permit is not required by Radiological Controls personnel during the performance of their assigned radiation protection duties provided they are following radiological control procedures for entry into High Radiation Areas.
6-22 Amendment Nos.
11, 35, 72, 77,106,107,12g,173
6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 GPU Nuclear Corporation initiated changes to the PCP:
1.
Shall be submitted to the NRC in the Annual Radioactive Effluent Release Report for the period in which the changes were made. This submittal shall contain:
a.
sufficiently detailed information to justify the changes without benefit of additional or supplemental information; b.
a determination that the changes did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and, c.
documentation that the changes have been reviewed and approved pursuant to 6.8.2.
2.
Shall become effective upon review and approval by GPUNC Management.
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6-23 Amendment Nos.
JJ, 75, 77, 77,195,197,129,173 i
4 6.14 0FFSITE DOSE CALCULATION KANUAL (0DCM) 6.14.1 The 00CM shall be approved by the Commission prior to implementation.
6.14.2 GPU Nuclear Corporation initiated changes to the ODCM:
1.
Shall be submitted to the NRC in the Annual Radioactive Effluent Release Report for the period in which the changes were made. This submittal shall contain:
a.
sufficiently detailed information to justify the changes without benefit of additional or supplemental information; b.
a determination that the changes did not reduce the accuracy or reliability of dose calculations or setpoint determinations; and i
c.
documentation that the changes have been reviewed and approved pursuant to 6.8.2.
2.
Shall become effective upon review and approval by GPUNC Management.
i 6.15 DELETED 6.16 POST-ACCIDENT SAMPLING PROGRAMS NUREG 0737 (II.B.3. II.F.1.2)
Program which will ensure the capability to accurately sample and analyze vital areas under accident conditions have been implemented.
The following programs have been established:
1.
Iodine and Particulate Sampling 2.
Containment Atmosphere Sampling Each program shall be maintained and shall include the following:
1.
Training of personnel, 2.
Procedures, and 3.
Provisions for maintenance of sampling and analysis equipment.
6-24 Amendment No.
7E, 77,102,173
t 6.17 MA).Q3 CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS 6.17.1 GPU Nuclear Corporation initiated safety related changes to the radioactive waste system (liquid, gaseous and solid):
i 1.
Shall be reported to the Commission in the Annual Report (Specification 6.9.1B) for the period in which the avaluation was reviewed. The discussion of each change shall contain:
i a.
A summary of the evaluation that led to the detemination that the change could be made in accordance with 10 CFR 50.59 b.
Sufficient detailed infomation to totally support the reason for the change without benefit of additional or supplemental information; c.
A detailed description of the equipment, components and processes involved and the interfaces with other plant-systems;-
d.
An evaluation of the change which shows.the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from i
those previously predicted in the license application and amendments thereto, i
e.
An evaluation of the change which shows the expected maximum exposures to individuals in the unrestricted area and to the general populatf on that differ from those previously estimated in the license application and i
amendments thereto;
[
f.
A comparison of the predicted releases of radioactive j
materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made; i
i g.
An estimate of the exposure to' plant operating personnel as a result of the change;.and h.
Dccumentation of the fact that the change was reviewed andapproved.
2.
Shall become effective upon review and approval in accordance with Section 6.5.1.
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l 6-25 i
Amendment No. 72, 77, 102,173
,