ML20137E756
| ML20137E756 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 08/31/1985 |
| From: | CONNECTICUT YANKEE ATOMIC POWER CO. |
| To: | |
| Shared Package | |
| ML20137E754 | List: |
| References | |
| GL-83-43, NUDOCS 8508260055 | |
| Download: ML20137E756 (27) | |
Text
a.
Docket 50-213 Attachment i Haddam Neck Plant Proposed Revisions to Technical Specification August,1985 8508260055 8506153 PDR ADOCK 0500021 PDR P
PAGE SECTION 2.0 SAFETY LIMITS AND MAXIMUM SAFETY SETTINGS C
2.1 Introduction 2.2 Reactor Core 2-1 2.3 Reactor Coolant System Pressure 2-1 2.4 Protective Instrumentation 2-4 2-5 SECTIC': 3.0 LIF.!TI.'.3 CONDITIONS FOR OPERATIO."
3.1 Introduction 3-1 3.2 Reactor Coolant System Activity 3-2 3.3 Reactor Coolant System Operational Components 3-3 3.4 Combined Heat-up, Cooldown and Pressure Limitations 3-5 3.5 Chemical and Volume Control System 3.6 Core Cooling Systems 3-8 3-10 3.7 Minimum Water Volume and Baron Concentration in the Refueling Water Storage Tank 3.8 Turbine Cycle 3-11 3-12 3.9 Operational Safety Instrumentation and Control C
Systems 3-14 3.10 Control Group ' Insertion Limits 3.11 Containment 3-16 3.12 Station Service Power 3-18 3.13 Refueling 3-21 3-23 3.14 Primary System Leakage 3.15 Reactivity Anomalies 3-25 3-27 3.16 Isothermal Coefficient of Reactivity 3-28 3.17 Limiting Linear Heat. Generation Rate 3-30 3.18 Power Distribution Monitoring and Control
- ~3.19 Hydraulic Snubbers 3-32 3-38 3.20 Reactor Coolant System Flow, Temperature and Pressure 3-39 3.21 Safety Related Equipment Flood Protection 3.22 Fire Protection Systems 3-40a 3.23 Post Accident Instrumentation 3-41 3-45 l
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{ SECTION 4.0 SURVEILLAN;E REQUIREMENTS PC2E 4.1 introduction Ahl 4.2 Operational Safety Items 4h2 4.3 Core Cooling Systems - Ppriodic Testing 4h3 4.4 Containment Testing th5 4.5 Emergency Power System Periodic Testing 4hil 4.6 Radioactive Liquid Waste Release 4b13 4.7 Radioactive Gaseous Waste Release 4h15 4.8 Auxiliary Steam Generator Feed Pump 4kl7 4.9 Main Steam Isolation Valves 4h18 4.10 Inservice Inspection and Reactor Vessel Surveillance 4'19 4.11 Deleted t.12 'High Energy Piping System Tests 4-24 4.13 Hydraulic Snubbers 4L26 4.14 Flood Protection Annunciation 4h29a.
4.15 Fire Protection Systems 4h30 4.16 Post Accident Instrumentation 4-33 l
SECTION 5.0 DE51GN FEATURES
- 5. l' Introduction 5-1 5.2 Site Description 5-1 5.3 Reactor 562 5.4 Containment 5-4 4
9 9
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III
INDEX
. DEFINITIONS '
SECTION 1.0 DEFINITIONS PAGE 1.1 Defined Terms 1-1 1.2 Thermal Power 1-1 1.3 Rated Thermal Power 1-1 1
1.4 Operational Mode 1-1 1.5 1-1 1.6 Operable - Operability 1-1 1.7 Reportable Event 1-2 1.8 Containment Integrity 1-2 1.9 Channel Calibration 1-2 1.10 Channel Check 1-2 1.11 Channel Functional Test 1-3 1.12 Core Alteration 1-3 1.13 Shutdown Margin 1-3 1.14 Identified Leakage 1-3 1.15 Unidentified Leakage 1-3 1.16 Pressure Boundary Leakage 1-4 1.17 Controlled Leakage 1-4 1.18 Quadrant Power Tilt Ratio 1-4 1.19 l-4 1.20 1-4 i
1.21 Frequency Notation 1-4 1.22 1-5 1.23 1-5 1.24 Flux Difference 1-5 1.25 Lower Power Physics Tests 1-5 Table 1.1 Operational Modes 1-6 Table 1.2 Frequency Notation 1-7
.t-
INDEX ADMINISTR ATIVE CONTROLS SECTION 6.0 ADMINISTRATIVE CONTROLS
_PAGE 6.5.2.4 Consultants 6-9 6.5.2.5 Meeting Frequency 6-9 6.5.2.6 Quorum 6-9 6.5.2.7 Review 6-10 6.5.2.8 Audits 6-11 6.5.2.9 Authority 6-11 6.5.2.10 Records 6-12 6.5.2.11
- Qualifications 6-12 6.6 REPORTABLE EVENT ACTION 6-13 6.7
. SAFETY LIMIT VIOLATION 6-13 6.8 PROCEDURES 6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS 6-15 6.9.2 SPECIAL REPORTS 6-18 6.10 RECORD RETENTION 6-18 6.11 RADIATION PROTECTION PROGRAM 6-20 6.12 RESPIRATORY PROTECTION PROGRAM (DELETED) 6.13 HIGH RADIATION AREA 6-20 6.14 ENVIRONMENTAL QUALIFICATION 6-21 6.15 SYSTEMS INTEGRITY 6-21 6.16 IODINE MONITORING 6-21 Haddam Neck v
DEFINITIONS 1.7 A reportable event shall be any of those conditions specified in Section 50.73 to 10CFR Part 50.
CONTAINMENT INTEGRITY
- 1.8 CONTAINMENT INTEGRITY shall exist when:
1.8.1 The containment automatic isolation valve system is OPERABLE.
1.8.2 All penetrations required to be closed during accident conditions are either:
a.
Capable of being closed by OPERABLE containment automatic isolation valves, or b.
Closed by manual valves, blind flanges, or deactivated automatic valves secured and locked in their closed positions, 1.8.3 ' All equipment hatches are closed and sealed, and 1.8.4 Each door in each air Im!: is closed and sealed except when the air lock is being used for normal transmit entry and exit through the containment, then at least one air lock door shall be closed and sealed.
CHANNEL CAllBR ATION 1.9 ' A CHANNEL CALIBRATION shall be the adjustment of channel output such that is responds, with acceptable range and accuracy, to known values of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip.
CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication with other indications
- derived from independent instrument channels measuring the same parameter.
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3.22 FIRE PROTECTION SYSTEMS Applicability This specification applies to' the operational status of the plant fire protection and detection systems.
Objective To ensure the availability of the fire protection and detection systems.
Specification A.
Fire Suppression Water Systems A.1 The Fire Suppression Water System shall be operable during all modes of operation with:
a)
Two high pressure fire pumps, each with a capacity of 2500 gpm, with their discharge aligned to the fire suppression header.
b)
Automatic initiation logic for each fire pump.
A.2 From and af ter the date that one fire pump is determined to be inoperable, restore the pump to operable status within 7 days or prepare and submit a Report to the Commission pursuant to Specification 6.6.1.
A.3 From and af ter the time that the entire Fire Suppression Water System is determined to inoperable with the plant in any operating mode:
a)
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> establish a backup Fire Suppression Water System, and prepare and submit a Report pursuant to Specification 6.6.1.
b)
The plant shall be placed in at least Hot Standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Cold Shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Haddam Neck 3-41
B.
Carbon Dioxide Systems B.1 The'following CO2 Systems shall be Operable with the minimum number of bottles of CO2 indicated, connected and available for service and having at least 90% of full charge weight, whenever the protected equipment in the area is required to be Operable:
a)
Cable Vault (18) b)
Primary Auxiliary Building Charcoal Filters (8)
B.2 From and af ter the time a CO2 system listed in B.1 is determined to be inoperable, within one hour establish is continuous fire watch with backup fire suppression equipment for the unprotected equipment and/or area.
B.3 Restore the system to Operable status within 14 days or prepare and submit a Report to the Commission pursuant to Specification 6.6.1.
C.
Halon 1301 System C.! The Halon 1301 System for the switchgear room shall be operable with at least (7) Halon bottles connected and available for service and each bottle shall be at least 92% of full charge weight.
' C.2 From and af ter the time the Halon system listed in C.1 is determined to be inoperable within one hour establish a continuous fire watch with backup fire su equipment for the unprotected equipment and/or area.ppression C.3 Restore the system to operable status within 14 days or prepare and submit a Report to the Commission pursuant to Specification 6.6.1.
D.
Fire Water Stations D.1 The Fire Water Stations listed in Table 3.22-1 shall be Operable.
Haddam Neck 3-42
I D.2 From and af ter the time any of the Fire Water Stations listed in Table 3.22-1 are determined to be inoperable, within one hour route an additional equivalent capacity hose to the unprotected area from an operable fire water station or establish a fire patrol to inspect the unprotected area at intervals of at least once each hour.
E.
Fire Detection Systems E.1 The minimum number of detectors for each fire detection zone shown in Table 3.22-2 shall be operable.
E.2 ACTION Whenever the number of OPERABLE detectors is less than the minimum number OPERABLE requirements of Table 3.22-2:
a.
Within I hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect the containment at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or (monitor the containment air temperature at least once per hour at the locations referenced by Specification 4.4, Containment i
Testing),
b.
Restore the inoperable instrument (s) to OPERABLE status within 14 days, or, prepare and submit a Report to the Commission pursuant to Specification 6.6.1.
F.
Penetration Fire Barriers F.1 Penetration fire barriers (including cable penetration barriers, fire doors and fire dampers) that protect safety related areas shall be functional.
F.2 From and af ter the time that it is determined that a penetration fire barrier is not functional; (a) the areas shall be monitored by operable fire detectors and within one (1) hour a fire watch patrol shall be established to inspect the area of at least once per hour or (b) a continuous fire watch shall be established on at least one side of the affected penetration within one (1) hour.
Haddam Neck 3-43
G.2 From and af ter the time any of the-spray and/or sprinkler systems listed in C.1 is determined to be inoperable, within one hour, establish a continuous fire watch with backup fire suppression for those areas starred (*) in C.1; for other areas, establish a fire patrol to inspect the applicable unprotected area at intervals of at least once each hour.
G.3 Restore the inoperable system to operable status within 14 days or prepare and submit a Report to the Commission pursuant to Specification 6.6.1.
Basis A.
Fire Detection and Suppression Systems Fire Suppression Systems The operability of the fire suppression systems ensures that adequate fire suppression capability is available to confine and extinguish fires occurring in any portion of the facility where safety related equipment is located.
The fire suppression system consists of the water systems, spray and/or sprinklers, CO2 Systems, Halon 1301 Systems, fire hose stations and hydrants. The capability of the fire suppression systems is adequate to minimize potential damage to safety related equipment and is a major element in the facility fire protection program.
In the event that a portion of the Fire Suppression System becomes inoperable, alternate backup fire fighting equipment or a periodic fire watch patrol is required to be established in the associated area until the inoperable equipment can be restored to service.
In the event that the fire suppression water system becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant.
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Haddam Neck 3-44a l
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- 3.23 Post Accadene Instrumentation Applicability This specification applies to specific plant in st rument at ion required for-post-accident assessments.
Obj ect ive To ensure.the instrumentation will be operable if required.
Speeification A.
Containment Hich Radation Monitor A.1 Two containment high radiation monitor channels shall be operable during modes 1, 2, 3, and 4 with:
a)
The capability to measure up to 1 x 108 R/hr.
b)
The alarm setpoint less than or equal 100 R/hr.
A.2 W1th one or more channel (s) inoperable, restore the inoperabic channel (s) to operable status within 7 days, or else prepare and submit a Special Report to the Commission pursuant to
(
Specification 6.9.3 within the next 10 days outlining: 'the cause of the malfunction, the plans for restoring the channel to operable status, and any alternative methods for esti-mating containment dose rates during the interim.
L 3-45
B.
_ Stack Wide Range Noble Gas Monitor C.
One stack vide range noble gas monitor channel shall be B.1 operable during modes 1, 2, 3, and 4 with:
a)
The capability to measure up.to 1 x-105 uC/cc of Xe-133.
b)
The alarm setpoint less than or equal to the monitor reading corresponding to I uC/cc.
B.2 With less than one-channel operable, return one channel to operable status within 7 days, or else prepare and submit a Special Report to the Commission pursuant Specification 6.9.3 within the next 10 days outlining:
to the cause of the malfunction, the plans for restoring the channel to operable status, and any alternative-methods for estimating stack release rates during the interim.
Basis These specifications ensure the operability of the high range containment radiation monitors and wide range stack noble gas monitors.
These monitors are required for post-accident assess-ments.
This capability is consistent with the recommendations of Regulatory Guide 1.97, " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions
-(
During and Following and Accident",
" Clarification of TMI Action Plan Requirements," NovemberDecember 1980 a
, 1980.
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- 4.10 IllSERVICE INSPECTICN AND REACTCR VESSEL SURVEILLANC
'_ App M cability:
and testing oT ASME Boiler and Pressure Yessel Code C C
1, 2, and 3 equivalent cosponents.
Objective:
applicable temponents defined above.To verify the structural integr Specification:
A.
Inservice Inspection of ASME Code Class 1 Class 2 and Class 3 equivalent components shall be performed
.in accordarice with Section XI' of the A5ME Boiler
, and Pressure Vessel Coce and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), with the exemptions and alternate inspections that have been approved by the NRC pursuant to 10 CFR 50, Section 10.55a(g)(6)(i). These exemptions and alternate inspections are included in the Inservice Inspection Program.
B.
Inservice testing of ASME Code Class 1. Class 2, and Class 3 equivalent pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(c), with the exemptions and alternate testing that have been
'(
50.55a(g) y the NRC pursuant to 10 CFR 50, Section approved b
~
(6) (i). These exemptions and alternate test,ing are. included in the Inservice Inspectier. Program.
C.
One reactor coolant pump flywheel shall be examined visually and 100* voluestrically every other refueling.
LA:
The removal plan for the reraining reactor vessel surveillance capsules is scheduled as follows:
Capsule Capsule Removal Time Predicted Excesure (E-1MEV)N d
' Type Jdent. _'
Years'
.Caosuie heactor Vessel L3 It g
Standby 1
B Standby 1
C Standby 1
5 Standby L
k 4-19
Basis 10 CFR 50.55a(g) sets forth the inspection requirements for nuclear power plant components by referencing.the ASME Boiler and Pre.ssure Vessel Code,Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components" which defines the requirements for inspection of ASME Code Class 1, 2, and 3, equivalent components.
There is one capsule (type II) containing specimens of wcld metal remaining (Capsule E).
In addition there are three type I capsules remaining (Capsules B, C, and G).
All four remaining capsules will remain in standby for special purposes.
4-20
4.16 Post Accident Instrumentation A.
Containment Hich Radiation Monitor A.1 Each containment high radiation monitor channel shall be demonstrated operable by:
a)
Performance of a channel check once per day.
b)
Perforrance of a channel calibration and functional test once every refueling.
Calibration of the sensor with a radioactive source need only be performed for levels less than or equal to 10R/hr. Higher levels may be calibrated electronically.
B.
Stack k'ide Range Noble Gas Monitor B.1 The stack wide range noble gas monitor channel shall be demonstrated operable by:
a)
Performance of a channel check once per day.
b)
Performance of a channel calibration and functional test once every 92 days.
Basis
(
The test frequencies specified will ensure the operability of the containment and stack monitors.
a.
m 0
L 4-33
(
CONNECTICUT YANKEE ATOMIC PO'a'ER COMPANY OFFSITE ORGANIZATION CYAPCO Chairman NUSCO Chairman and Chief Executive Officer CYPACO President NUSCO President and Chief Operating Officer CYAPCO Executive Vice President and Chief Operating Officer s
NUSCO Executive Vice President Engineering and Operations
(
CYAPCO Senior Vice President NUSCO Senior Vice President Nuclear Engineering and Operations l
I Nuclear Review Northeast utilities Board Service Company (NUSCO)*
CYAPCO Vice President NUSCO Vice President Nuclear Operations Nuclear Operations Connecticut Yankee Station Staff Superintendent - Onsite
- Provides Operating and Technical Suppor,t by Contractual Arrangement Tigure 6.2-1 HADDAM NECK 6-2
ADMINISTRATIVE CONTROLS
(
6.5.2 NUCLEAR REVIEW BOARD (NRB)
FUNCTION 6.5.2.1 The Nuclear Review Board shall function to provide independent review and audit of designated activities in the areas of:
J
- a. nuclear power plant operations
- b. nuclear engineering
- c. chemistry and radiochemistry
- d. metallurgy t
- e. nondestructive testing
- f. instrumentation and control
- g. radiological safety
- h. mechanical and electrical engineering
- 1. administration
- j. quality assurance practices 9
6-8 L
4
ADMINISTRATIVE CONTROLS 4
COMPOSITION 6.5.2.2 The NRB shall consist of no less than seven, nor more than eleven members including the Chairman and the Station Superintendent.
The Chairman and members of the NRB shall be app ~ointed in writing by the Senior Vice President - Nuclear Engineering and Operations, and shall have an academic degree in engineering or phy,sical science field or equivalent ex-perience.
Technical ~ experience in the areas specified in Section 6.5.2.1 may be substituted on a one-for-one basis towards the academic degree requirement.
(Four years of technical experience being equivalent to a four-year academic degree.)
In addition, they shall have a minimum of five years technical experience, of which a minimum of three years shall be in their respective field of expertise.
CONSULTANTS 6.5.2.3 Consultants shall be utilized as determined by the NRB Chairman to provide expert advice to the NRB.
l MEETING FREOUENCY 6.5.2.4 The NRB shall meet at least once per six months.
l QUORUM 6.5.2.5 The minimum quorum of the NRB necessary for the performance of the NRB review and audit functions of these Technical Specifications l shall consist of the Chairman or his designated alternate and at least four-(4) NRB members.
No more than a minority of the quorum shall have line responsibility for operation of the facility.
L 6-9
ADMINISTR ATIVE CONTROLS REVIEW 6.5.2.6 The NRB shall review:
The safety evaluations for 1) changes to procedures, equipment or a.
systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
b.
Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59,10 CFR.
Proposed tests or experiments which involve an unreviewed safety c.
question as defined in Section 50.59,10 CFR.
d.
Proposed changed in Technical Specifications or licenses.
e.
Violations of applicable statutes, codes, - regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance, f.
Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety.
g.
A!! reportable events.
h..
Indications of a significant unanticipated deficiency, affecting nuclear safety, in some aspect of design or operation of safety related structures, systems or components.
1.
Reports and meeting notes of the PORC.
Haddam Neck 6-10
f ADMINISTR ATIVE CONTROLS 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for Reportable Events:
a.
The Commission shall be notified and/or a report submitted pursuant to the requirements of Section 50.73 to 10CFR Part 50 and b.
Each Reportable event shall be reviewed by-the PORC and submitted to the NRB and the Vice President Nuclear Operations. -
6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is
- violated, The unit shall be placed in at least HOT STANDBY within one hour.
a.
b.
The NRC. Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Vice President Nuclear Operations and the NRB shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
A Safety Limit Violation Report shall be prepared. The report c.
shall be reviewed by the PORC. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon f acility components, systems or structures, and (3) corrective action taken to prevent recurrence.
d.
The Safety Limit Violation Report shall be submitted to the Commission, the NRB and the Vice President-Nuclear Operations within 14 days of the violation.
6.8 PROCEDURES 6.8.1 Written procedures and administrative policies shall be established, implemented and maintained that meet or exceed the requirements and recommendations of Section 5.1 and 5.3 of ANSI N18.7-1976 and Appendix "A" of USAEC Regulatory Guide 1.33 except as provided in 6.8.2 and 6.8.3 below. Procedures shall be established and maintained for implementation of the Facility Fire Protection Program.
6.8.2 Each procedure and administrative policy of 6.8.1 above, and changes thereto, shall be reviewed by the PORC and approved by the Station Superintendent prior to implementation and periodically as set forth in each document.
Haddam Neck 6-13
s 6.9 REPORTING REQUIREMENTS in addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the regional administrator of the appropriate. Regional Office unless otherwise noted.
6.9.1 Routine Reports Startup Report - A summary report of plant startup and a.
power escalation testing shall be submitted following (1) receipt of an operating license,- (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal,' or hydraulic. performance of the plant.
The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. ' Any corrective actions that were required to obtain satisfactory operation shall be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.
Startup-reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption.
or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest.
If the Startup Report does not cover all three. events (i.e.,
initial criticality, completion of-startup test program, and resumption or commencement of commercial power operation), supplementary reports shall be submitted at i.
least every three months until all three events have been completed.
b.
Steam Generator Tube Inspection - The complete results of the steam generator tube inservice inspection shall be reported within ninety days after completion of the inspection. This report shallinclude:
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 plugged.
Haddam Neck 6-15 t
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c.
Occupational Exposure Report The annual report shall be submitted prior to March I of each year and shall cover the previous calendar year. The initial report shall be submitted prior to March I 'of the year following initial criticality.
This an.nual report shall include a tabulation on an annual basis of the number of staiion, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /yr. and their associated man-rem exposure according to work and job functions, 4 e.g., reactor operations and surveillance; inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing and refueling.
The dose assignment to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements.
Small exposures totalling less than 20% of the individual total does need not be accounted for. In the aggregate, at least 80% of the total wnole body dose receive from external sources shall be assigned to specific major work functions.
d.
Monthly Operating Report - Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Management Information and Program Control, U. S. Nuclear Regulatory Commission, Washington, D. C.
20555, with a copy to the appropriate Regional Office, to be postmarked no later than the fif teenth of each month following the calendar month covered by the report.
The monthly report shall include a narrative summary of operating experience during the report period relating to safe operation of the facility, including safety-related maintenance, e.
Reports required as per 10 CFR 50.59b.
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ADMINISTRATIVE CONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the appropriate regional office within the time period specified for each report.
.These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference, specification:
- a. Inservice inspection results, specifications (4.10) and (4.12).
- b. Primary Containment Leak Rate Results, Specification (4.4).
- c. Reactor Vessel Material Surveillance Specimen Examination, Specification (4.10).
- d. Steam Generator Tube Report.
Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission within 15 days.
6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:
a.
Records and logs of facility operation covering time interval at each power level.
b.
Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
c.
All Reportable _ Events.
d.
Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
5 l
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ADMINISTR ATIVE CONTROLS Records or reactor tests and experiments.
e.
f.
Records of changes made to Operating Procedures.
g.
Records of radioactive shipments.
h.
Records of sealed source leak tests and results.
i.
Records of annual physical inventory of all source material of record.
j.
Radiation and contamination surveys.
6.10.2 The following records shall be retained for the duration of the Facility Operating License:
a.
Record and drawing changes reflecting facility design modifications made to systems and equipment described in the Facility Description and Safety Analysis Report.
- b.. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
Records of facility radiat' ion and contamination surveys, c.
d.
Records of radiation exposure for all individuals entering radiation control areas.
Records and gasecus and liquid radioactive material released to the e.
- environs, f.
Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles.
Records of training and qualification for current members of the g.
plant staff.
h.
Records of in-service inspections performed pursuant to these Tbchnical Specifications.
i.
Records of Quality Assurance activities required by the QA
- Manual, j.
Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.
k.
Records of meetings of the PORC and the NRB.
1.
Records for Environmental Qualification which are covered under the provisions of paragraph 6.14.
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ADMINISTR ATIVE CONTROLS 6.11 R ADIATION 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 SUPERSEDED BY 10 CFR 20.103(c) AND (f) ON DECEMBER 23,1977 6.13 HIGH R ADIATION AREA 6.13.1 in lieu of the " control device" or " alarm. signal" required by paragraph 20.203(c)(2), each high radiation area in which the intensity of radiation is 1000 mrem /hr or less shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Exposure Authorization *.
An - individual or group of individuals permitted to enter such areas shall be provided with one or more of the following:
A radiation monitoring device which continuously indicates the a.
radiation dose rate in the area.
b.
A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the. area have been established and personnel have been made knowledgeable of them, A health physics qualified individual (i.e. qualified in radiation c.
protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and who will perform periodic radiation surveillance at the frequency specified in the REA.
The surveillance frequency will be established by the Health Physics Supervisor.
The above procedure shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintaine~d under the administrative control of the shif t supervisor on duty.
- Health Physics personnel shall be exempt from the REA issuance requirement during the performance of their assigned radiation protection duties, providing they are following plant radiation areas.
radiation protection procedures for entry into high Haddam Neck 6-20
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ADMINISTR ATIVE CONTROLS 6.14 ENVIRONMENTAL QUALIFICATION A.
By no later than June 30,.1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of Division of Operating Reactors " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 " Interim Staff Position on Environmental -Qualification of Safety-Related Electrical Equipment", December 1979. Copies of these documents are attached to Order the Modification of License-DPR-61, dated October 24,1980.
B.
By no later than December 1,1980, complete and auditable records must be available and maintained at a central location which describe the environmental qualification method used f.or a!!
safety-related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588.
Therefore, such ' records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.
6.15 SYSTEMS INTEGRITY The licensee shall implement a program to reduce leakage from systems outside containment that would and could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.
This program shall include the following:
- 1. Provisions establishing preventive maintenance and periodic visualinspection requirements, and
- 2. Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.
6.16 10 DINE MONITORING The licensee shall implement a program which will ensure the capability to accurately determine the airborne iodine concentration in vital area under accident conditions. This program shall include the following:
- 1. Training of personnel,
- 2. Procedure for monitoring, and
- 3. Provisions for maintenance of sampling and analysis equipment.
Haddam Neck 6-21 I.