ML20214N628
| ML20214N628 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/11/1986 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20214N149 | List: |
| References | |
| GL-83-43, NUDOCS 8609160393 | |
| Download: ML20214N628 (15) | |
Text
._
I.
TECHNICAL SPECIFICATION CHANGE REQUEST (TSCR) NO. 138, SUPPLEMENTAL 1 2
The licensee requests that the following page changes be made to supplement TSCR 138, dated tiay 12, 1986 (5211-86-2079).
a.
Replace the following pages:
111, iv, v, 3-107, 3-109, 3-112, 3-113, 4-81, 4-85, 6-13 b.
Add the following pages:
4-56 and 6-14 c.
Delete the following page: 4-57 i
II.
REASON FOR CHANGE The purpose of TSCR 138 was to revise reporting requirments consistent with rule changes to 10 CFR 50.72 and 50.73 as requested by Generic Letter (G.L.) 83-43 and also to incorporate other administrative changes affecting those same pages as described in TSCR 138.
Subsequent to GPUN submittal of TSCR 138 it has been determined that certain modifications to TSCR 138 are either necessary or desirable.
The purpose of this supplement is to modify TSCR 138 to include those additional changes.
III.
SAFETY EVALUATION JUSTIFYING CHANGE i
The changes incorporated by this supplement are administrative in nature and serve to clarify the existing technical specifications and include additional elements of the Standard Technical Specifications (STS).
The following is a descripticn and justification for the changes included in this supplement:
a.
Specification 4.19.5 is being modified to require notification in l
accordance with 10 CFR 50.72 prior to resumption of plant operation in the event of steam generator tube inspection results which f all into Category C-3.
Standard Technical Specifications and also the current TMI-l Technical Specifications require that this report be made prior to resumption of plant operation. The words "... prior to resumption of plant operation" were deleted in TSCR 138 since it was felt that it would not be possible for the plant to resume l
operation prior to the required notification. The words "... prior to resumption of plant operation" are being incluted at the request of the NRC in order to reflect the Standard Technical Specification wordi ng.
b.
Specification 4.19.5 is also being modified to allow 12 months for submittal of the complete results of steam generator tube inservice inspections. This change is consistent with Standard Technical Specifications. The requirement for submittal of this report within 3 months following completion of the inspection was incorporated as part of Technical Specification Amendment No. 47 ( n December 22, 1978, which was prior to the issuance of the STS tr a.1 allows 12 months for submittal of the report.
c.
The proposed Bases for specification 3.22.1.3 is being modified to delete the last sentence. This sentence is not part of the STS Bases, but was included in order to reference an example of where the NRC had employed a fractional limit similar to the way the proposed Tech. Spec. employed a fraction. Since the reference may not be clear, GPUN wishes to delete it from the proposed amendment.
d.
Consistent with Standard Technical Specifications, the bases for Specification 3.22.1.2 has been revised to reference the Safe Drinking Water Act, 40 CFR 141.
This Act is not applicable to R11-1 release limits, however it is used as a reference, for comparative purposes only, in the required report.
e.
Other changes that are proposed as part of this supplement are editorial in nature and are included in order to improve the punctuation or format.
This supplement reflects changes to TMI-1 Technical Specifications received subsequent to submittal of TSCR 138, up through Amendment No. 120.
IV.
NO SIGNIFICANT HAZARDS CONSIDERATION GPU Nuclear Corporation has determined that this Technical Specification Change Request Supplement poses no significant hazards as defined by NRC regulations in 10 CFR 50.92.
This ensures that 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 (10 CFR 50.92(c)(1)) since the proposed amendment does not involve plant design or operational changes; or 2.
create the possibility of a new or different kind of accident from any accident previously evaluated (10 CFR 50.92(c)(2)) since the proposed amendment does not involve modifications to existing plant equipment; or.
~
3.
involve a significant reduction in a margin of safety since the proposed amendment does not involve changes to the conditions assumed in the accident analyses (10 CFR 50.92(c)(3)).
The Commission has provided guidelines pertaining to the application of the three standards by listing specific examples in 48 FR 14870. The proposed amendment to incorporate TSCR 138 and this supplement is considered to be in the same category as the following examples of amendments that are considered not likely to involve significant hazards consideration:
(i)
A purely administrative change to achieve consistency, correct errors, change nomenclature and improve clarity, and (vii) A change to make a license conform to changes in the regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations.
The proposed amendment does not affect plant design or operation, does not involve modification of plant equipment, does not involve changes to any plant release limits, and does not affect the plant's safety a nalysi s.
Therefore, the three standards of 10 CFR 50.92(c) are satisfied.
In summary, GPU Nuclear Corporation has determined and submits that the proposed administrative changes described herein do not represent any significant hazards.
V.
IMPLEMENTATION tle request that the amendment approving this TSCR be made effective upon issuance.
VI.
AMENDMENT CLASSIFICATION (10 CFR 170.2)
TSCR 138 was submitted in part at the request of the Commission. NRC Generic Letter 83-43 stated that remittance of a license fee for these changes was not required.
To cover those portions of TSCR 138 which were not specifically related to Generic Letter 83-43, pursuant to the provisions of 10 CFR 170.21, a check for $150.00 was included with GPUN's submittal of TSCR 138, dated May 12, 1986. This supplement should be considered as part of TSCR 138.
Therefore an additional application fee is not required. _
TABLE OF CONTENTS Section Page 3.16 SHDCK SUPPRESSORS (SNUBBERS) 3-63 3.17 REACTOR BUILDING AIR TEMPERATURE 3-80 3.18 FIRE PROTECTION 3-86 3.18.1 FIPE DETECTION INSTRUMENTATION 3-86 3.18.2 FIRE SUPPRESSION WATER SYSTEM 3-88 3.18.3 DELUGE / SPRINKLER SYSTEMS 3-89 3.18.4 CO2 SYSTEM 3-90 3.18.5 HALON SYSTEMS 3-91 3.18.6 FIRE HOSE STATIONS 3-92 3.18.7 FIRE BARRIER PENETRATION SEALS 3-94 3.19 CONTAINMENT SYSTEMS 3-95 3.20 SPECIAL TEST EXCEPTIONS 3-95a 3.20.1 LOW POWER NATURAL CIRCULATION TEST 3-95a i
3.21.1 RADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION 3-96 1
3.21.2 RADI0 ACTIVE GASEOUS PROCESS AND EFFLUENT 3-100 MONITORING INSTRUMENTATION 3.22 RADIOACTIVE EFFLUENTS 3-106 3.22.1 LIQUID LtFLUENTS 3-106 3.22.2 GASE0US EFFLUENTS 3-111 3.22.3 SOLID RADI0 ACTIVE WASTE 3-118 3.22.4 TOTAL DOSE 3-119 3.23 RADIOLOGICAL ENVIRONMENTAL MONITORING 3-120 3.23.1 MONITORING PROGRAM 3-120 3.23.2 LAND USE CENSUS 3-125 3.23.3 INTERLABORATORY COMPARISON PROGRAM 3-127 4
SURVEILLANCE STANDARDS 4-1 1
4.1 OPERATIONAL SAFETY REVIEW 4-1 4.2 REACTOR COOLANT SYSTEM INSERVICE INSPECTION 4-11 4.3 TE5ilNG FOLLOWING OPENING OF SYSTEM 4-28 4.4 REACTOR BUILDING 4-29 4.4.1 CONTAINMENT LEAKAGE TESTS 4-29 4.4.2 STRUCTURAL INTEGP.ITY 4-35 4.4.3 DELETED 4-37 2
4.5 EMERGENCY LOADING SEQUENCE AND POWER TRANSFER, 4-39 EMERGENCY CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4.5.1 EMERGENCY LOADING SEQUENCE 4-39 4.5.2 EMERGENCY CORE COOLING SYSTEM 4-41 4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM 4-43 4.5.4 DECAY HEAT REMOVAL SYSTEM LEAKAGE 4-45 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTS 4-46 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4-48 4.7.1 CONTROL ROD DRIVE SYSTEM FUNCTIONAL TESTS 4-48 4.7.2 CONTROL R00 PROGRAM VERIFICATION 4-50
-iii-Amendment 72, 81, 108
TABLE OF CONTENTS Section Page 4.8 MAIN STEAM ISOLATION VALVES 4-51 4.9 DECAY HEAT REMOVAL CAPABILITY - PEPIODIC TESTING 4-52 4.9.1 EMERGENCY FEEDWATER SYSTEM - PERIODIC TESTING 4-52 (REACTOR COOLANT TEMPERATURE GREATER THAN 250*F) 4.9.2 DECAY HEAT REMOVAL CAPABILITY - PERIODIC TESTING 4-5?a (REACTOR COOLANT TEMPERATURE 250*F OR LESS) 4.10 REACTIVITY AN0iALIES 4-53 4.11 REACTOR COOLANT SYSTEM VENTS 4-54 4.12 AIR TREATMENT SYSTEMS 4-55 4.12.1 EMERGENCY CONTROL ROOM AIR TREATMENT SYSTEM 4-55 4.12.2 REACTOR BUILDING PURGE AIR TREATMENT SYSTEM 4-55b 4.12.3 AUXILIARY & FUEL HANDLING EXHAUST AIR TREATMENT SYSTEM 4-55d 4.13 RADI0 ACTIVE MATERIALS SOURCES SURVEILLANCE 4-56 4.14 DELETED 4-56 l
4.15 MAIN STEAM SYSTEM INSERVICE INSPECTION 4-58 4.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE 4-59 4.17 SHOCK SUPPRESSORS (SNUBBERS) 4-60 4.18 FIRE PROTECTION SYSTEMS 4-72 4.18.1 FIRE PROTECTION INSTRUMENTS 4-72 4.18.2 FIRE SUPPRESSION WATER SYSTEM 4-73 4.18.3 DELUGE / SPRINKLER SYSTEM 4-74 4.18.4 CO2 SYSTEM 4-74 4.18.5 HALON SYSTEMS 4-75 4.18.6 HOSE STATIONS 4-76 4.18.7 FIRE BARRIER PENETRATION SEALS 4-76a l
4.19 OTSG TUBE INSERVICE INSPECTION 4-77 4.19.1 STEAM GENERATOR SAMPLE SELECTION & INSPECTION METHODS 4-77 4.19.2 STEAM GENERATOR TUBE SAMPLE SELECTION & INSPECTION 4-77 4.!9.3 INSPECTION FPEQUENCIES 4-79 4.19.4 ACCEPTANCE CRITERIA 4-80 4.19.5 REPORTS 4-81 4.20 PEACTOR BUILDING AIR TEMPERATURE 4-86
[
4.21.1 PADI0 ACTIVE LIQUID EFFLUENT INSTRUMENTATION 4-87 4.21.2 RADI0 ACTIVE GASEOUS PROCESS & EFFLUENT MONITORING 4-90 l
INSTRUMENTATION l
4.22.1 LIQUID EFFLUENTS 4-97 1
4.22.2 GASEOUS EFFLUENTS 4-105 4.22.3 SOLID RADI0 ACTIVE WASTE 4-115 l
4.22.4 TOTAL DOSE 4-116 4.23.1 MONITORING PROGRAM 4-117 4.23.2 LAND USE CENSUS 4 -1 21 l
l 4.23.3 INTERLABORATORY C0'iPARISON PROGRAM 4-122
-iv-Amendment No. 11, 28, 30, 41, 47, 55, 72, 78, 95, 97, 119 L
TABLE OF CONTENTS Section Page 5
DESIGN FEATURES 5-1 5.1 SITE 5-1 5.2 MITAINMENT 5-2 5.2.1 REACTOR BUILDING 5-2 5.2.2 REACTOR BUILDING ISOLATION SYSTU4 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 QUALIFICATIONS 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 6-11 6.9 REPORTING REQUIREMENTS 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 SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT 6-18 6.10 RECORD RETENTION 6-19 6.11 RADI ATION PROTECTION PROGRAM 6-21 6.12 HIGH RADI ATION AREA 6-21 6.13 PROCESS CONTROL PROGRAM 6-21 6.14 0FF5ITE DOSE CALCULATION MANUAL (0D01) 6-22 6.15 DELETED 6-22 6.16 POST ACCIDENT SAMPLING PROGRAMS 6-22 NUREG 0737 (II.B.3, II.F.1.2) 6.17 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS 6-23
-v -
Amendment No, 11, 47, 72, 77
RADIOACTIVE EFFLUENTS LIQUID EFFLUENTS DOSE LIMITING CONDITION FOR OPERATION 3.22.1.2 The dose or dose commitment to an individual from radioactive materials in liquid effluents released from the unit to the site boundary (see Figure 5-4) shall be limited:
a.
During any calendar quarter to < 1.5 mrom to the total body and to < 5 mrem to aliy organ, and I
b.
During any calendar year to < 3 mrem to the total body and to < 10 mrem to any organ.
APPLICABILITY:
At all times ACTION:
a.
With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the subsequent 3 calendar quarters so that the cumulative dose or dose commitment to any individual from such releases during these four calendar quarters is within 3 mrem to the total body and 10 mrem to any organ. This Special Report shall also include (1) the result of radiological analyses of the drinking water source, and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Water Act.
BASES This specification is provided to implement the requirements of Sections II.A, III. A, and IV. A of Appendix I,10 CFR Part 50.
The Limiting Condition for Operation implements the guides set forth in Section II. A of Appendix I.
The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive material in liquid effluents will be kept "as low as is reasonably achievable". Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess l
of the requirements of 40 CFR 141.
The dose calculations in the ODCM implement l
3-107 Amendment No. 72 j
RADI0 ACTIVE EFFLUENTS LIQUID RADWASTE TREATMENT SYSTEM LIMITING CONDITION FOR OPERATION 3.22.1.3 The appropriate portions of the liquid radwaste treatment system l
shall be used to reduce the radioactive materials in ifquid wastes prior to their discharge when the projected doses due to the liquid effluent from the unit to unrestricted areas (see Figure 5-4) would exceed 0.06 mrem to the total body or 0.2 mrem to any organ in any calendar month.
APPLICABILITY: At all tims ACTION:
a.
With radioactive liquid waste being discharged without treatment and in excess of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which includes the following information:
1.
Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for inoperability, 2.
Action (s) taken to restore the inoperable equipment to OPERABLE status, and
[
3.
A summary description of action (s) taken to prevent a l
recurrence.
BASES The requirement that the appropriate portions of this system be used, when specified, provides assurance that the releases of radioactive materials in liquid effluents will be kept as low as is reasonably achievable. This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section II.D of Appendix I to 10 CFR Part 50. The intent of Section II.D. is to reduce effluents to as low as is reasonably achievable in a cost effective manner. This LCO satisfies this intent by establishing a dose limit which is a small fraction (25%) of Section II.A of Appendix I, 10 CFR Part 50 dose requirements.
This margin, a factor of 4, constitutes a reasonable reduction.
3-109 Amendment No. 72
RADI0 ACTIVE EFFLUENTS GASEOUS EFFLUENTS DOSE-NOBLE GASES LIMITING CONDITION FOR OPERATION 3.22.2.2 The air dose due to noble gases released in gaseous effluents from the unit to areas at and beyond the site boundary (see Figure 5-3) shall be limited to the following:
a.
During any calendar quarter: < 5 mrad for gamma radiation and < 10 mrad for beta radiation, and l
b.
During any calendar year: < 10 mrad for gamma radiation and
< 20 mrad for beta radiatioii.
APPLICABILITY: At all times.
ACTION:
a.
With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
BASES This specification applies to the release of radioactive materials in gaseous effluents from TMI-1.
This specification is provided to implement the requirements of Section II.B.
!!I. A and IV. A of Appendix I,10 CFR Part 50.
The limiting Cond' fon for Operation implements the guides set forth in Section II.B of Appendix I.
The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure
(
that the releases of radioactive material in gaseous effluents will be kept "as low as is reasonably achievable." The Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Release of Reactor 4
3-112 Amendment No. 72 l
RADI0 ACTIVE EFFLUENTS GASEOUS EFFLUENTS DOSE - 10 DINE-131,10 DINE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE FORM LIMITING CONDITION FOR OPERATION 3.22.2.3 The dose to an individual from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents released from the unit to areas at and beyond the site boundary (See Figure 5-3) shall be limited to the following:
a.
During any calendar quarter: j 7.5 mrem to any organ, and b.
During any calendar year: j 15 mrem to any organ.
APPLICABILITY: At all times.
ACTION:
With the calculated dose from the release of iodine-131, iodine-133, tritium, and radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the NRC Region I Administrator within 30 days, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.
BASES This specification applies to the release of radioactive materials in gaseous effluents from THI-1.
This specification is provided to implement the requirements of Section II.C, III. A and IV. A of Appendix I,10 CFR Part 50.
The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix 1.
The ACTION statement provides the required operating flexibility and at the same time implements the guides set forth in Section IV. A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III. A of Appendix I that ccnformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.
The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of 3-113 Amendment 72
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4.13 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE Applicability Appifes to leakage testing of byproduct, source, and special nuclear radioactive material sources.
Objective To assure that leakage from byproduct, source, and special nuclear radioactive material sources does not exceed allowable limits.
Specification Tests for leakage and/or contamination shall be performed by the licensee or by other persons specifically authorized by the Commission or an agreement State, as follows:
1.
Each sealed source, except startup sources previously subject to core flux, containing radioactive material, other than Hydrogen 3, with a half-life greater than 30 days and in any form other than gas shall be tested for leakage and/or contamination at intervals not to exceed six months.
2.
The periodic leak test required does not apply to sealed sources that are stored and not being used.
The sources excepted from this test shall be tested for leakage prior to any use or transfer to another user unless they have been leak tested within six months prior to the date of use or transfer.
In the absence of a certificate from a transferor indicating that a test has been made within six months prior to the transfer, sealed sources shall not be put into use until tested.
3.
Each scaled source shall be tested within 31 days prior to being subjected to core flux and following repair or maintenance to the source.
i i
I l
4.14 DELETED l
l l
4-56 (page 4-57 deleted) l Amendment No. 64 3 81 L
a
b.
The steam generator shall be determined OPERABLE af ter completing the corresponding actions (removal from service by plugging, or repair by the kinetic expansion process, of all tubes exceeding the repair limit and all tubes containing throughwall cracks) required by Table 4.19.2.
4.19.5 Reports a.
Following the completion of each inservice inspection of steam generator tubes, the number of tubes repaired or removed from service in each steam generator shall be reported to the NRC within 15 days.
b.
The complete results of the steam generator tube inservice inspection shall be reported to'the NRC within 12 months l
following completion of the inspection. This report shall I
include:
1.
Number and extent of tubes inspected.
2.
Location and percent of wall-thickness penetration for each indication of an imperfection.
3.
Identification of tubes repaired or removed from service.
c.
Results of steam generator tube inspections which fall into Category C-3 require notification in accordance with 10 CFR 50.72 prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine the cause of the tube degradation and corrective measures taken to prevent recurrence in accordance with 10 CFR 50.73.
Bases The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be
[
maintained, i
l
{
l l
l l
i l
l 4-81 Amendment No. 47, 83, 91, 103 L
i 1
1 F
3 TABLE 4.19-2 i
STEAM GENERATION TUBE INSPECTIONI2)
'd 1
II II l
5 l
IST SAMPLE INSPECTION ll 2ND SAMPLE INSPECTION l l 3RD SAMPLE INSPECTION l
I l l l l l
[
t Sample Size l Result I Action Required l l Result I Action Required l l Result I Action Required l
1 I
I ll 1
l l 1
l l A a:inicum of l C-1 l
None ll N/A l
N/A 1l N/A I
N/A l
15 Tu l
C-2 i Plug or repair ll C-1 i
None l l N/A l
N/A l
15.G.ge}per Il l
1 defective tubes l l C-2 i Plug or repair l l C-l I
None l
l l
1 and inspect l l l defective tubes and l l l
l 1
I l additional 25 ll l inspect additional 4S I l C-2 i Plug or repair l
l l
l tubes in this ll l tubes in this S.G.
l l l defective tubes.
l l
l l S.G.
ll l Perform action for ll l Perform action l
1 l
l ll C-3 l C-3 result of first l l C-3 I for C-3 result l
l 1
1 Il l sample.
l l I of first sample.
l l
l C-3 I Inspect all l l Other l
l l l
l l
l l tubes in this l I S.G. is l None l l N/A N/A 1
l l
l S.G., plug or l l C-1 l
l l l
l 1
I l repair defect-l l Other 1 Perform action for l l l
l l
l ive tubes and l l S.G. is l C-2 result of second l l N/A I
N/A i
1 l
l inspect 25 tubesi l C-2 l sample l l l
l p
l l
1 in other S.G.
l l Other l Inspect all tubes in l l l
l 3
l l
l Provide notifi-l l S.G. is I each S.G. and plug or l l N/A I
N/A l
l l
1 cation to NRC l l C-3 i repair defective l l l
l l
l l pursuant to ll l tubes. Provide l l l
l l
l l 10CFR50.72.b.2.il l l notification to NRC l l l
l l
l l and submit a l l l pursuant to 10CFR50.72l l l
l l
1 I report pursuant l l l b.2.i and submit a l l 1
l l
l l to 10CFR50.73.- l l l report pursuant to l l l
l 1
l l a.2.ii.
l 1 l 10CFR50.73.a.2. ii.
l I I
l Notes:
(1) S=3 E % Where N is the number of steam generators in the unit, and n is the number of steam generators inspected during an inspection.
(2) For tubes inspected pursuant to 4.19.2.a.4:
No action is required for C-1 results. For C-2 results in one or both steam generators plug or repair defective tubes. For C-3 results in one or both steam generators, plug or repair defective tubes and provide notification to NRC pursuant to 10 CFR 50.72.b.2.1 followed by a written report pursuant to 10 CFR 50.73.a.2.ii.
b
2.
The following information on aircraf t movements at the Harrisburg International Airport:
a.
The total number of aircraf t movements (takeoffs and landings) at the Harrisburg International Airport for the previous twelve-month period, b.
The total number of movements of aircraf t larger than 200,000 pounds at the liarrisburg International Airport for the previous twelve-month period, broken down into scheduled and non-scheduled (including military) takeoffs and landings, based on a current estimate provided by the airport manager or his designee.
3.
The following information from the periodic Leak Reduction Program tests shall be reported:
a.
Results of leakage measurements, b.
Results of visual inspections, and c.
Maintenance undertaken as a result of Leakage Reduction Program tests or inspections.
4.
The following information regarding pressurizer power operated relief valve and pressurizer safety valve challenges shall be reported:
a.
Date and time of incident, b.
Description of occurrence, and c.
Corrective measures taken if incident resulted from an equipment failure.
5 The following information regarding the results of specfic activity analysis in which the primary coolant exceeded limits of Technical Specification 3.1.4.1 shall be reported:
a.
Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; b.
Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiofodine activity was reduced to less than limit.
Each result should include date and time of sampling and the radiofodine concentrations; c.
Cleanup system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; d.
Graph of the I-131 concentration and one other radiciodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and e.
The time duration when the specific activity of the primary coolant exceeded the radiofodine limit.
l 6-13 Amendment No. 11, 37, 72, 77, 82, 117 L
4 C.
Monthly 6perating Reports. Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the U.S.
Nuclear Regulatory Commission, no later than the fif teenth of each month l
following the calendar month covered by the report.
6.9.2 DELETED 6-14 (Pages 6-15 and 6-16 deleted)
Amendment No,11, 77