ML20005A122
| ML20005A122 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/24/1981 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20005A120 | List: |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 NUDOCS 8106290416 | |
| Download: ML20005A122 (13) | |
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l ATTACHMENT II PROPOSED TECHNICAL. SPECIFICATIONS CHANGES RELATED TO NUREG-0578 TMI-2 LESSONS LEARNED CATEGORY "A" ITEMS POWER AUTHORITY OF THE STATE OF.NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 I
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TABLE 3.2-6 -
SURVEILLANCE INSTRUMENTATION' Miminum No.
of Operable No..of Channels Instrument Type Indication Provided
. Channels- Instrument 'and Range by Design Action 1:
(Reactor Level' Indicator) (13) (2) l
( -(Note 3) 0 - +60 ) .5-2 ( ) O '
(Reactor. Level Recorder )
( (Note 4) 0 - +60 )
1 Reactor' Level Indicator
-150 - +60 2 (2)
(Reactor Pressere Indicator )
( (Note 5) 0-1200 psig) .
2 ( ) 5 (1) (2)
(Reactor Pressure Recorder )
( (Note 6) 0-1200 psig)
(Drywell-Pressure (Narrow Range)
( (Narrow Range) Indicator )
( Recorder )
( 10 - 19 psia )
1 ( ) 2 (2).
(Drywell Fressure (Wide Range) )
.( (Wide Range) Indicator )
( Recorder -)
( 0 - 100 psia )
(Drywell Temperature Indicator )
( 50 - 2500 F )
2 ( ) 4 (1) (2)
(Drywell Temperature' Recorder. )
50 - 3500F )
i (Suppression Chamber Indicator )
(Temperature H50 - 2500 F~-)
I
'2 ( . . .) 4 (1) (2)
(Suppression Chamber Recorder )
(Temperature 50 - 3500 F .)
Amendment No. J$ , 4S , 76
TABLE 3.2-6 ,
SURVEILLANCE INSTRUMENTATION
' Minimum No.
cf Operable No. of Channels Instrument Type Indication Provided Channels Instrument and Range by Design Action, (Suppression Chamber Indicator )
(Water Level Recorder )
( (Wide Range) -72 to + 72 inches) 1 ( ) 2 (2)
(Suppression Chamber Indicator )
(Water Level Recorder )
( (Narrow Range) -6 to +6 inches )
N/A Control Rod Indicator 1 (7)
Position Indication Fosition 00 to 48 2 Source Range Indicator 4 (8)
Monitors Recorder f
1 to 10' cps 3 Intermediate Indicator 8 (8) (9)
Range Monitor Recorder 10-4 to 40% Rated Power 2 Average Power Indicator 6 (8) (9)
Range Monitor Recorder 0-125% Rated Power 1 Drywell-Suppression Recorder 2 ;2) i Chamber Differential O to 5 psi
! Presrure Computer 0 to 5 psi 1 Safety / Relief Valve Indicator 2 (12) (11)
Position Indicator Open/ Closed (Note 10)
NOTE 3 FOR TABLE 3.2-6
- 1. From and after the date that the minimum number of operable instrument channels is one less than the minimum number specified for each parameter, continued operation is permissible during the succeeding 30 days unless the minimum number specified is made operable sooner.
Amendment No. 76a
NOTES FOR TABLE 3.2-6 (CONTINUED)
- 2. In the ev?nt thct all indications of this parameter is disabled and such indication cannot be .
restored in six (6) hours, an orderly shutdown shall be initiated and the reactor shall be. in a Hot Shutdown condition in six (6) hours and a Cold Shutdown condition in the following -
eighteen (18) hours.
- 3. Three (3) indicators from level instrument channel A, B, & C. Channel A or B are utilized-for feedwater control, reactor water high and low level alarms, recirculation pump-runback. High level trip of main turbine and feedwater pump turbine utilizes chtnnel A, B, & C.
- 4. One (1) recorder utilized the same level instrument channel as selected for feedwater control.
- 5. Three (3) indicators from reactor pressure instrument channel A, B, & C. Channel A cr B are utilized for feedwater control and reactor pressure high a' arm.
- 6. One (1) recorder. Utilizes the same reactor pressure instrument channel'as selected'for feedwater control.
- 7. The posi tion of each of the 137 control rods is monitored by the Rod Position Information System. For control rods in which the position is unknown, refer to Paragraph 3.3.A.
- 8. Neutron monitoring operability requirements are specified by Table 3.1-1 and Paragraph 3.3.B.4.
- 9. A minimum of 3 IRM or 2 APRM channels respectively must be operable (or tripped) in each safety system.
- 10. Each Safety Relief Valve is equipped with two primary acoustical detectors (of which one is in service). A thermocouple detector serves as a secondary indicator.
- 11. From and after the date that none of the acoustical detectors is operable but the thermocouple is operable, continued operation is permissible until the next outage in which a primary containment entry is made. Both acoustical detectors shall be made operable prior to restart.
- 12. In the event that both primary and secondary indications of this parameter for any one valve are disabled and neither indication can be restored in forty-eight (48) hours, an orderly shutdown shall be initiated and the reactor shall be in a flot Shutdown condition in twelve (12) hours and in a Cold Shutdown within the next twenty-four (24) hours.
- 13. From and after the date that the minimum number of operable instrument channels is one less than the minimum number specified for each parameter, continued operation is permissable during the succeeding 7 days unless the minimum number specified is made operabic sooner.
Amendment No. 49 76b
TABLE 4.2-6 ,
MINIMUM TEST AND CALIBRATION FREQUENCY FOR SURVEILLIANCE INSTRUMENTATION IMSTRUMENT CHANNEL CALIBRATION FREQUENCY INSTRUMEfff CHECK 1.) Reactor Water Level Once/6 months Once Each Shift 2.) Reactor Presi:ure Once/6 months Once Each Shift 3.) Drywell Pressure Once/6 months. Once Each Shift 4.) Drywell Temperature Once/6 months Once Each Shift 5.) Suppression Chamber Temperature once/6 months Once Each Shift 6.) Suppression Chamber Water Level Once/6 months Once Each Shift 7.) Control Rod Position Indication N/A Once Each Shift 8.) Neutron Monitoring (APRM) Five/ week Once Each Shift 9.) Neutron Monitoring (IRM and SRM) Note 10 Note 10 10.) Drywell-Suppression Chamber Differential Pressure Once/6 months Once Each Shift 11.) Safety / Relief Valve Position Indicator (Primary) Note 11 Once/ Month 12.) Safety / Relief Valve Note'll Once/ Month Position Indication (Secondary) .
Amendment No. [, f4f 84
NOTES FOR TABLES 4.2-1 '1HROUGH 4.2-6
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- 1. Initially once every month until acceptance failure rate ' 8. Uses same instrumentation as Main Steam data are available thereafter, a request may be made to the Line High Radiation. See Table 4.1-2.
l NRC to change the test frequency. The compilation of instrumene failure rate data may include data obtained from 9. See Technical Specification 1.0.F.4, other boiling water reactors for which the same design Definitions, for meaning of term, instrument operate in a environment similar to that of " Instrument Check".
- 10. Calibration and instrument check surveillance
- 2. Functional tests, calibrations and instrument checks are not for SRM and IRM Instruments are as specified required when these instruments are not regoired to be in Table 4.1-1, 4.1-2, 4.2-3.
operable or are tripped. Functional tests shall be performed before each startup with a required frequency not 11. Functional test is performed once each to exceed once per week. Calibrations shall be performed operating cycle.
prior to each startup or prior to preplanned shutdowns with a required frequency not to exceed once per week.
Instrument checks shall be performed at least once per day during these periods when the instruments are required to be operable.
- 3. This instrumentation is excepted from the functional test definition. The functional test will consist of injecting a simulated electrical signal into the measurement channel.
These instrument channels will be calibrated using simulated electrical signals once every three months.
- 4. Simulated automatic actuation shall be perforned once each .
operating cycle. Where possible, all logic system functional tests will be performed using the test jacks.
- 5. Reactor low water level, high drywell pressure and high radiation main steam line tunnel are not included on Table 4.2-1 since they are tested on Table 4.1-2.
- 6. The logic system functional tests shall include a calibration of time delay relays and timers necessary for proper functioning of the trip systems.
- 7. At least one (1) Main Stack Dilution Fan is required to be in operation in order to isokinetically sample the Main .
Stack.
Amendment No. Jd', JP 85
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- 6. In cddition to itema 1, 2 & 3 tbova, two cuditionnl cperatoro chcIl be rondily cvoiltblo en cito whsnsvar the reactor is in other than cold shutdown. During cold shutdown, an additional operator shall be readily available on site.
- 7. An individual qualified in radiation protection proce-dures shall be on site when fuel is in the reactor.
- 8. In the event of illness or absenteeism up to two (2) hours is allowed to restore the shift crew or fire-brigade to normal complement.
- 9. A Fire Brigade of five (5) or more members shall be maintained on site at all times. This excludes two (2) members of the minimum shift crew necessary for safe shutdown and any personnel required for other essential functions during a fire emergency.
- 10. A Shift Technical Advisor shall be on site and readily available to the control room except during the cold shutdown or refuel mode.
6.3 PLANT STAFF QUALIFICATIONS .
The minimum qualifications with regard to educational background and experience for plant staff positions shown in Fig. 6.2-1 shall meet or exceed the minimum qualifications of ANSI NI8.1-1971 for comparable positions; except for the Radiation and Environmental Services Superintendent who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 and the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents. Any deviations will be justified to the NRC prior to an individual's filling of one of these positions.
6.4 RETRAINING AND REPLACEMENT TRAINING A training program shall be maintained under the direction of the Training Coordinator to assure overall proficiency of the plant staff organization. It shall consist of both retraining and replacement training and shall meet or exceed the minimum requirements of Section 5.5 of ANSI NI8. 1-1971.
The retraining program shall not exceed periods two years in length with a curriculum designed to meet or exceed the requalification requirements of 10 CFR 55, Appendix A. In addition fire brigade training shall meet or exceed the require-ments of NFPA 27-1975, except for Fire Brigade training sessions which shall be held at least quarterly. The effective date for implementation of fire brigade training is March 17, 1978.
6.5 REVIEW AND AUDIT l Two seperate review groups for the review and audit of plant l operations have been constituted. One of these, the Plant The l
Operating Review Ccmmittee (PORC), is an onsite group.
other is an independent review and audit group, the offsite Safety Review Committee (SRC).
248 f knendment No. jpg, yf'
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- a 6.5.1 PLANT OPERATING REVIEW COMMITTEE (PORC)
(A) Membership The PORC is comprised of the Resident Manager (Chairman) , Superintendent of Power (Vice Chairman),
Operations Superintendent, Maintenance Superintendent, Technical Services Superintendent, Instrument and Control Superintendent, Radiological and Environmental Services Superintendent and Reactor Analyst. Special consultant to provide expert advice may be utilized when the nature of a particular problem dictates.
248a i
Amendment No.)('
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(B) Alternates Alternate members shall be appointed in writing by the PORC Chairman to serve on a temporary basis; however, no more than two alternates shall participate in PORC activities at any one time.
(C) Meeting Trequency Meetings will be called by the Chairman as the occasions for review or' investigation arise. Meetings will be no less frequent than once a month.
(D) Quorum The Chairman or Vice Chairman and four members, including designated alternates, shall constitute a quorum.
(E) Rcsponsibilities
- 1. Review plant procedures, and changes thereto, re-quired by Specification 6.2.
- 2. Review proposed tests and experiments that affect nuclear safety.
- 3. Review proposed changes to the Operating License and Technical Specifications.
- 4. Review proposed changes or modifications to plant systems or equipment that affect nuclear safety
, 5. Investigate violations of the Technical Specifications and prepare and forward a report covering evaluation '
and recommendations to prevent recurrence to the Resident Manager, who will forward the report to the Manager - Nuclear Operations and to the Chairman of the Safety Review Committee.
- 6. Review plant operations to detect potential safety hazards.
- 7. Review the Security Plan and implementing procedures annually.
249 kiendment No. jHF
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' S t r o -- st riroa arAcroa openATon POWER AUTHORITY OF THE STATE OF NEW YORK l.0 - I:1 AL.'To h Ol'E N A TO!t JAMES A. FITZPATRICK NUCLEAR POWER PLANT PLANT STAFF ORGANIZATION
l-4' ATTACHMENT III SAFETY EVALUATION OF CHANGES RELATED TO NUREG-0578 TMI-2 LESSONS LEARNED CATEGORY "A" ITEMS 4
e POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 1
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Section I - Description of Modification The James A. FitzPatrick Operating License is being amended by adding license conditions related to a System Integrity Measurements Program and Improved Iodine Measurements capability. Technical Specification Sections 3.2 (Table 3.2-6) and 4.2 (Table 4.2-6) are i being changed to incorporate limiting conditions for operation and surveillance requirements for the Safety Relief Valve position indicators. Technical Specification Section 6.2 and 6.3 (including Figure 6.2-1) were revised to reflect the addition of a Shift Techaical Advisor to the plant staff. These changes were requested by the NRC staff in a letter dated July 2, 1980. Figure 6.2-1 has also been corrected to more accurately reflect the Reactor Analyst's position i.e., much greater involvement with operations personnel than with Technical Services personnel.
Section II - Purpose of Modification The purpose of this modification is to incorporate the items requirements into the*,
of the NUREG-0578 TMI-2 Lessons Learned Category "A" James A. FitzPatrick Operating License and Technical Specifications as requested in the NRC letter of July 2, 1980.
Section III - Impact of the Change These modifications will not alter the conclusions reached in the FSAR and SER acc.ident analysis.
Section IV - Implementation of the Modification The modifications as proposed will not impact the ALARA or Fire Protection Program at JAF.
Section V - Conclusion The incorporation of there modifications: a) will not increase the probability nor the consequences of an accident or mal-function of equipment important to safety as previously evalu-ated in the Safety Analysis Report; b) will nut increase the possibility for an accident or malfunction of a different typeand than any evaluated previously in the Safety Analysis Report; c) will not reduce the margin of safety as defined in the basis for any Technical Specification, and d) does not constitute an unreviewed safety question.
Section VI - References (a) JAF FSAR (b) JAF SER wmmm - ,_ - _