ML20058N013
| ML20058N013 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 12/10/1993 |
| From: | Boulette E BOSTON EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20058N016 | List: |
| References | |
| BECO-93-156, NUDOCS 9312210194 | |
| Download: ML20058N013 (17) | |
Text
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BOSTON EDISON Pilgnm Nuclear Power Station Rocky Hill Road Piymouth, Massachusetts 02360 10 CFR 50.90 E. T. Boulette, PhD Senior Vce President- Nuclear U. S. Nuclear Regulatory Commission BECo 93 156 Document Control Desk December 10, 1993 Washington, DC 20555 License DPR-35 Docket 50-293
Subject:
Proposed Chanae to lechnical Specifications:
Reauest for Chanaes Supportina a 24 Month Fuel Cycle (Submittal 3)
Boston Edison Company (BECo) proposes the attached Technical Specification Change Request to Operating License No. DPR-35 in accordance with 10CFR 50.90. This is the last of the three change request submittals supporting the adoption of a 24 month refueling cycle at Pilgrim.
Information supporting this change request is contained in Attachment A to this letter, the proposed replacement pages are contained in Attachment B, and marked-up pages are contained in Attachment C. A consolidated version of the replacement pages from all 3 submittals i< provided as Attachment D. An Instrument Summary by Technical Specification Secti:h is provided in Attachment E.
Boston Edison requests NRC approval of this amendment bv March,1994, or earlier to support our planning of Mid-Cycle Outage #10, currently scheduled to begin in the Fall of 1994.
h[JO L E.T. Boulette, PhD Senior Vice President - Nuclear Commonwealth of Massachusetts)
County of Plymouth )
Then personally appeared before me, E. T. Boulette, who being duly sworn, did state that he is Senior Vice President - Nuclear of Boston Edison Company and that he is duly authorized to execute and file the submittal contained herein in the name and on behalf of Boston Edison Company and that the statements in said submittal are true to the best of his knowledge and belief.
My commission expires: dod/tt.
DATE 47O NOTARY PUBL m0%3 9312210194 931210 PDR ADOCK 05000293 Ii 00I P PDR Il1 M
Boston Edison Company U.S. Nuclear Regulatory Commission Page 2 Attachments: (A) Description of Proposed Change (B) Amended Technical Specification Pages (C) Marked-up Pages from Current Technical Specifications (D) Consolidated Copy of Amended Technical Specification Pages (E) Instrument Summary by Technical Specification Section 1 signed original and 37 copies ETB/JDK/nas/24MONTHC cc: Mr. R. Eaton, Project Manager Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Mail Stop: 1401 U. S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville. MD 20852 U. S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Mr. Robert M. Hallisey, Director Radiation Control Program Center for Communicable Diseases Mass. Dept. of Public Health 305 South Street Jamaica Plain, MA 02130 Senior NRC Resident Inspector Pilgrim Nuclear Power Plant
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Attachment A to BECo Letter 93-A. Description of Chanae This Technical Specification change will allow BECo to extend existing plant surveillance intervals to 24 months from 18 months. Intervals of eighteen months are being changed to once-per-cycle; intervals currently 1 once-per-cycle will be considered to be 24 months as reflected in the '
I revised definition of " Surveillance Interval". This submittal (submittal #3) changes specific setpoints to accommodate a 24-month fuel cycle and provides a justification for extending the surveillance l interval for those components and systems that are not related to instrument setpoint changes. In cases where the surveillance could be l performed safely on-line and justification of an extended interval was l not developed, the "once/ cycle" is changed to the currently allowed 18 months. Several of the bases pages were also revised to delete ;
references to the trip setpoint values, as these values are now included ;
as the Limiting Conditions of Operation. j l
In conjunction with the setpoint evaluations, BEco has reviewed the effects of increased calibration intervals on instrument drift. The increased calibration interval will not adversely affect plant operations or safety. The process and methodology for determining 1 setpoints was presented in our first 24 month change request dated June i 7, 1993 and is consistent with Generic Letter (GL) 91-04.
Certain components and systems that are not related to instrument l setpoints are currently surveilled at a frequency of once-per-cycle, l once/18 months, or during refueling outages. For these, BEco assessed .
the effect on safety of the change in surveillance intervals to I accommodate a 24-month surveillance interval (plus 25% for a bounding interval of 30 months) and determined the effect on safety to be minimal or non-impacting. Additional reviews of historical maintenance and surveillance data for select systems were performed to confirm these conclusions. A summary of these reviews is provided for each system.
Proposed changes are identified by vertical bars in the margin of the affected page in Attachments B and D.
Another element of GL 91-04 called for a summary description of the program for monitoring and assessit, the effects of increased calibration surveillance intervals on instrument drift and its effect on sa fety. Our Setpoint Control Prooram was developed for implementation in three phases. Phase 1 developed the methodology and controls used to l perform setpoint calculations and derive the setpoint values that were subsequently reflected in the three technical specification change submittal s . Specific details of the methodologies and controls of the program including a sample calculation were provided in our first submittal.
The second phase of the program w:.s established to address the design change mechanism to be used for effecting setpoint changes and to address the associated procedural controls to be used to maintain control of the critical parameters identified in the calculations.
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These parameters typically include: technical specification allowable j values, field setpoints, instrument reset values, M&TE requirements, no adjust limits, and channel cross check values. These parameters are identified in the "Results and Conclusions" section of the calculations and are considered design basis data. As such, setpoint changes will be controlled in accordance with our plant design change process. The setpoint design changes and associated surveillance procedure revisions will be made before the technical specification surveillance intervals l are extended. The calculations also identified the total loop uncertainty and the margin associated with the calculated allowable value and the field setpoint. l l
The third phase of the program consists of the monitoring, feedback, and assessment to verify instrument performance, including drift, is ,
consistent with the parameters specified in the calculations. Data collected from surveillance procedures will be evaluated to confirm the assumptions supporting the setpoint calculations and the conclusions in the calculations are valid. The proposed (revised) setpoints will be evaluated during the 1st operating cycle to ensure the new setpoints have no adverse impact on plant operations. The extended surveillance ,
intervals will be monitored for three refueling cycles to ensure the l assumptions in the calculations continue to be valid. If surveillance i test results indicate that instrument performance does not meet surveillance procedure requirements, corrective actions will be taken in j accordance with existing station procedures. Changes made as a result :
of the corrective action will be reflected in the setpoint calculations. '
Subsequent monitoring will rely on existing plant procedures and controls to ensure continued safe operation of Pilgrim Station.
Also, due to editorial errors and additional changes, pages 46a, 61 and ,
68 of the first submittal (BEco Letter 93-072 dated June 7,1993) and ,
pages 27, 39, 40 and 47 of the second submittal (BECo letter 93-097 l dated August 9,1993) are being superseded in this third submittal. I l
B. Reason for Proposed Change l l
During the April 1993 Refueling Outage Boston Edison loaded sufficient i fuel for a 2 year operating cycle. The proposed changes are requested l to support changing the ftal cycle at Pilgrim from 18 to 24 months. l The proposed changes contained in this submittal result from analyses performed in accordance with Generic Letter 91-04. The analyses in some cases will require a setpoint change to ensure the instruments perform their function over the extended interval. These changes are indicated on the marked up pages of the proposed Technical Specifications.
Other, non-setpoint, changes are supported by the review and discussion summarized under each system.
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C. Safety Discussion Reactor Protection System (Scram) Instrumentation - l Mode Switch in Shutdown i TS Table 4.1.1, item 1 Page 30 TS 4.1 requires: Reactor protection system (RPS) instrumentation and ,
associated aevices that initiate a reactor scram are to be functionally !
tested and calibrated according to the minimum frequency intervals shown in Table 4.1.1. The reactor mode switch in the Shutdown position is one ,
functional test listed in Table 4.1.1 with a specified test interval of l once every refueling outage. ,
l This test verifies the Reactor Mode Switch Shutdown Scram and Scram Bypass logic associated with the RPS function properly. The Reactor Mode Switch scram function is not required to protect the fuel or nuclear boundaries. The RPS functions independently from the mode ,
switch. The RPS takes precedence over the. mode switch in the event of 1 an undetected mode switch failure; therefore, the RPS could and does 1 provide both automatic and manual scram capability. Based on the l independence and redundancy of RPS and the Reactor Mode Switch, extending to a 24 month testing interval will not impact the availability and reliability of the mode switch shutdown position j function. A historical review of maintenance and surveillance data j confirms this conclusion. ,
j Loaic System Functional Test (LSFT) j The following list of TS line items in the LSFT group are affected by changing from an 18 month to a 24 month fuel cycle:
l l TS Table 4.2. A items 1-5 Page 60 l TS Table 4.2.B items 1-9 Page 62 l
TS Table 4.2.C Page 63 TS Table 4.2.D items 1 and 2 Page 64 TS Table 4.2.G Page 66b TS 4.8.H Page 183 TS Table 4.2. A items 1-5, require: A LSFT be performed once/ 18 months i for the PCIS system, i l
TS Table 4.2.B, items 1-9 require: A LSFT be performed once/18 months !
for the Core Standby Cooling Systems (CSCS).
l TS Table 4.2.C requires: A LSFT vis-a-vis " System Logic Test" once/ cycle for Control Rod Block actuation.
TS Table 4.2.D, items 1 and 2 require: A LSFT be performed once/18 months for the Reactor Building Isolation and Standby Gas Treatment System (SGTS) Actuation.
TS Table 4.2.G requires: A LSFT be performed once per refueling cycle for the Recirculation Pump Trip.
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i TS 4.8.H requires for the Mechanical Vacuum Pump: "At least once during each operating cycle verify automatic securing and isolation of the mechanical vacuum pump".
The logic system functional test interval of 18 months was selected to minimize safety system inoperability due to testing and to minimize the potential for inadvertent safety system trips and their attendant transients. Performing LSFTs requires the installation of jumpers, lifting leads, blocking of contacts and/or bypassing safety functions.
This increases the potential for error and introduces an element of risk to the plant.
Because of the test complexity, conducting these tests on a once/ cycle frequency to coincide with the proposed 24 month fuel cycle will reduce the risk of inadvertently injecting into the vessel, isolating safety
! systems or overpressurizing low pressure systems. System unavailability and system outage time due to LSFT testing will also be reduced.
Standby safety system unavailability is a function of equipment failure rates, repair time and the test interval. The effect of an increased test interval is the increased likelihood of an undetected failure and a decrease in unavailability. Since logic system failure rates are significantly lower than those for mechanical components, safety system reliability is dominated by mechanical component failures. Accordingly, mechanical components are tested more frequently (quarterly) to ensure system reliability. Increasing the LSFT interval from 18 to 24 months will not affect the capability of these systems to perform their design function.
A review of the LSFT surveillance history supports this conclusion.
Simulated Automatic Actuations:
TS 4.5.A.1.a Page 103 TS 4.5.A.3.a Page 104 TS 4.5.C.I.a Page 107 TS 4.5.D.1.a Page 108 TS 4.5.E.1.a Page 109 TS 4.7.A.2.b.l.a Page 155a TS 4.5. A.1 requires: The Core Spray system have a Simulated Automatic Actuation test once/ operating cycle.
TS 4.5.A.3.a requires: The LPCI system have a Simulated Automatic Actuation test once/ operating cycle.
TS 4.5.C.1.a requires: The HPCI system have a Simulated Automatic Actuation test once/ operating cycle TS 4.5.D.1, item (a) requires: The RCIC system have a Simulated Automatic Actuation test performed once/ operating cycle.
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TS 4.5.E.1 requires: The Automatic Depressurization System have a Simulated Automatic Actuation test prior to startup from a refueling outage. l TS 4.7. A.2.b.l.a requires: The operable Primary Containment Isolation Valves that are power-operated and automatically initiated be tested for Simulated Automatic initiation and closure times at least once/ operating i cycle. 1 i
l The above referenced TS surveillances are performed prior to startup i following a refueling outage. The purpose of these requirements is to j ensure that the systems are tested and are available to perform their !
design function following a refueling outage. A review of the surveillance history for these items supports extending the surveillance :
interval to 24 months. Based on the review it is concluded that !
changing the surveillance interval to 24 months will not affect the capability of these systems to perform their design function. ;
Core Standby Coolina System (CSCS) Instrumentation Test and Calibration !
Frecuency T.S. Table 4.2.B Page 61 T.S. 4.2.B.7 requires: The Trip System Bus Power Monitor be functionally tested once/ operating cycle. l This requirement tests the CSCS power monitors (relays) capability to )
detect loss of voltage and annunciate in the control room. The relays !
l are normally energized and upon loss of voltage, de-energize and cause an annunciation in the Control Room. The functional test consists simply of pulling the fuse ahead of the CSCS relay to confirm the relay ;
de-energizes. A review of the relay survaillance history supports ;
extending the surveillance interval from 18 months to 24 months. ~
REACTIVITY CONTROL !
TS 4.3.A.1 Page 80 l l
TS 4.3. A.1 requires: Sufficient control rods be withdrawn following a :
refueling outage when core alterations were performed to demonstrate I with a margin of 0.25 percent Ak that the core can be made subcritical at any time in the subsequent fuel cycle with the strongest operable control rod fully withdrawn and all other operable rods fully inserted.
l The demonstration is made at the beginning of the cycle and the results I are adjusted to the most reactive condition in the cycle. Extending the cycle length, therefore, does not affect the validity of the demonstration. By withdrawing sufficient control rods following each refueling outage to demonstrate adequate margin exists, margin is assured through the cycle and is independent of cycle length.
CONTROL ROD DRIVES (CRD)
TS 4.3.B.l.a Page 81 TS 4.3.B.I.b Page 82 TS 4.3.C.1 Page 83 Page 5 of 15 i
TS 4.3.B.I.a requires: Control rod coupling integrity be verified when a rod is withdrawn for the first time after a refueling outage or after maintenance. TS 4.3.B.I.b requires control rod coupling integrity to be verifiei when a rod is fully withdrawn for the first time following a refueling or after maintenance.
1 Because these surveillance requirements are event dependent, increasing j the length of the operating cycle will have minimal effect on the capability of the CRD system to function as designed. Additional assurance of coupling integrity is provided throughout the cycle due to the performance of coupling checks anytime a control rod reaches the full out position.
6 TS 4.3.C.1 requires: The testing of scram insertion times after each refueling outage, or after a reactor shutdown greater than 120 days.
This requirement verifies that the control rod drive system is capable t
l of bringing the reactor subcritical at a rate sufficient to prevent fuel damage. The intent is to check for and predict potential failures j l
t associated with the CRD's after a refueling or shutdown of greater than ;
l 120 days. This requirement is based on event occurrence. Control rod !
inactivity over a period of 120 days is the event controlling the testing frequency. During RFOs when fuel is moved or reloaded into the core, the core geometry change warrants CRD testing.
During an l operating cycle, core geometry does not change and scram times should '
not vary for this reason. Increasing the fuel cycle from 18 to 24 months will have minimal safety impact on the CRD ability to function as ;
designed. Historical maintenance and surveillance data confirms this >
conclusion. Additionally, a qualitative test per T.S. 4.3.C. requires at least 10% of the control rods be scram timed every 120 days on a rotating basis. This plus a weekly exercise of control rods throughout ,
the cycle continue to provide assurance of system availability and function.
Standby Liouid Control System (SBLC) '
TS 4.4.C.4 Page 5'7 TS 4.4.C.4 requires: The SBLC solution (BIO) enrichment be checked by test anytime boren is added to the solution and during each refueling outage. ;
Enrichment is tested at an interval consirtent with the potential for it i to vary.
It will gt change unless Boron is added to the storage tank. .
The quantity of B stored ig the SLCS storage tank is sufficient to bring the concentration of B in the reactor to the point where the reactor will be shutdown and to provide a minimum shutdown margin at 25%. Extending the surveillance interval (18 months to 24 months) will not effect the availability and safety function of the SBLC.
HPCI and RCIC Flow Rate TS 4.5.C.I.e Page 107 ,
TS 4.5.D.1.e Page 108 i
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TS 4.5.C. I .e and 4.5.D.l.e require: A flow rate test at a reactor pressure of 150 psig once/ operating cycle for the High Pressure Core Injection (HPCI) pump and the Reactor Core Isolation Cooling (RCIC) l pump, respectively. l The purpose of these surveillances is to assure the HPCI and RCIC i Systems are operable upon startup from a refueling outage. The tests are performed at 150 psig to verify system operability at a low reactor pressure prior to proceeding to normal reactor operating pressure. The :
frequency is event dependent (i.e., startup from a RFO). In addition, I both the HPCI and RCIC systems, including the pumps, are subjected to j quarterly testing of various components. This testing monitors performance and provides early indication of degraded performance. The i 150 psig flow test is performed more frequently if the reactor is at or l below 150 psig and the appropriate monthly / quarterly tests are required.
The test is also performed if work has been performed on the system requiring a post-work demonstration of system operability by actually !
running the system. Historically, these tests are performed, on average, once/ year for reasons other than the once/ cycle requirement.
Therefore, based on the additional testing performed during the >
operating cycle and the once/ cycle test frequency being event dependent, !
increasing the allowed surveillance interval to 24 months will not impact on the HPCI or RCIC capability to function as designed. i Automatic Depressurization System (ADS)
TS 4.5.E.1.b Page 109 ;
TS 4.5.E.1.b requires: Manually opening each relief valve each operating cycle with the reactor at pressure to effect a change in reactor pressure or induce flow in the main turbine bypass valve.
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. This requirement tests the manual actuation of the ADS system and demonstrates each valve's ability to reduce reactor pressure. Other components comprising ADS are tested frequently throughout the operating cycle. The valve's location (inside primary containment) precludes inadvertent damage or tampering. Tailpipe temperature monitoring I
throughout the cycle detects potential valve degradation and provides added assurance that the relief valves are capable of performing their safety function. Increasing to a 24 month testing interval, therefore, will not impact on the capability of the relief valves to function as
! designed. A review of the maintenance and surveillance history confirms this conclusion.
l Leakaoe Detection Systems i l
! T.S.4.6.C.2.a.2 Page 125b T.S.4.6.C.2.b.3 Page 125b i
T.S. 4.6.C.2.a.2 requires: For each required drywell sump monitoring )
system, perform an instrument channel calibration at least once per 18 '
months.
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Nuclear system leakage rate limits are established so that appropriate action can be taken before the integrity of the nuclear system barrier is compromised. The drywell sump monitoring system consists of one equipment sump pump and one floor drain sump pump plus associated instrumentation. Flow integrators for each sump (equipment and floor) comprise the basic instrument system and are used to record the flow of l liquid from the sumps. A review of the surveillance history confirms that extending the interval to 24 months is justified. Additional technical specification required functional testing every 31 days provides assurance that the drywell sump monitoring system remains capable of performing its design function throughout the operating cycle.
T.S.4.6.C.2.b.3 requires: For each drywell atmospheric radioactivity monitoring system, perform an instrument channel calibration at least >
once per 18 months. ;
Nuclear System Leakage Rate Limits are established so that appropriate action can be taken before the integrity of the Nuclear System process barrier is compromised. The monitoring of airborne radioactivity levels in the containment atmosphere permits operators to evaluate leakage relative to the probable source (i.e., an abundance of iodine in the i containment atmosphere would indicate water leakage, whereas an abundance of gaseous activity would indicate steam leakage). Airbor;e radioactivity levels of the drywell atmosphere are monitored by the ,
Reactor Pressure Boundary Leak Detection System. This system consists of two panels, each with 3 channels, capable of monitoring the primary containment atmosphere. Performance of the monthly functional test also verifies the instrument channel is within calibration. Extending the minimum calibration frequency to once/ cycle (24 mos) will not reduce the monthly functional testing which in turn verifies the instrument channel is still within its calibration specifications. Based on the redundant design of the system and the total number of channels available (six), i and the monthly functional testing which will identify the need for i calibration if required during the operating cycle, assurance is l provided that the system will remain capable of performing its design 1 function.
Safety and Relief Valves TS 4.6.D.2 Page 126 TS.4.6.D.2 requires: At least one of the relief / safety valves be j disassembled and inspected each refueling outage. l This surveillance is event dependent and established to coincide with j refueling outages. Pl ant safety is not impacted, as the increased surveillance interval does not change the function or configuration of the safety and relief valves. A historical review of inspection and maintenance records confirmed that increasing the surveillance interval to 24 months will not affect the capability of the safety and relief valves to perform their safety function.
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i Shock Suporessors (Snubbers)
. TS 4.6.1 Page 137a e TS 4.6.I.2.A Page 137b
- TS 4.6.I.3.B Page 137d TS 4.6.1 o TJires
- A visual inspection of all safety related hydraulic and mechar.' snubbers listed in PNPS procedures every 18 months.
TS 4.6.I.2.A requires: A functional test of 10% of each type of snubber l at least once per operating cycle. j i
TS 4.6.I.3.B requires: A review of the installation and maintenance records for each safety related snubber listed in PNPS procedures once i per cycle. l The requirement to visually inspect snubbers once/ cycle was developed to allow inspection during outages sufficiently long to permit access to ,
snubbers inaccessible during power operation. Similarly, this interval j was selected to accommodate the need to functionally test snubbers that l were inaccessible during power operation, j 1
The snubber service life documentation is reviewed each cycle. If a snubber's service life would expire prior to the next scheduled review (next refueling outage), the snubber is either reconditioned, replaced 3 or its expected service life is re-evaluated. Performing preventive maintenance prior to the service life expiration provides assurance that snubbers will perform their design function. 'bn efore, an increase in the operating cycle to 24 months will have no impact on snubber service life. In order to maintain the same snubber population tested over a 10 year interval, we have increased the number of snubbers to be tested a each cycle. That change is reflectea in the revised specifications.
Suppression Chamber & Drvwell Surface TS 4.7.A.I.e Page 153 TS 4.7.A.2.d Page 155b TS 4.7. A.I.e requires: A visual inspection of the suppression chamber interior, including the water line regions, shall be made "at each major <
refueling outage". J TS 4.7. A.2.d requires: A visual inspection of the interior surface of the drywell and torus every refueling outage for evidence of deterioration.
1 This requirement addresses the degradation of the drywell and suppression chamber shells and other associated structural members due to corrosion.
I Pilgrim's drywell and suppression chamber are inerted with nitrogen '
during power operation, reducing corrosion rates. In addition, they are coated to prevent corrosion with a material demonstrated to have a long life span. A complete re-coating of the suppression chamber submerged Page 9 of 15
l portion was performed in 1982 due to extensive modifications made to the !
suppression chamber in that time frame. The upper portion above the -
water line and the drywell have the original coating with some touching l up in spots. Past surveillances indicate a low rate of corrosion !
induced degradation. j Therefore, based on the proven life of the coating material and on-inerting the drywell and suppression chamber, increasing the !
surveillance interval from 18 to 24 months will not result in ;
accelerated or undetected degradation of the drywell or suppression i chamber. l l
Containment Atmospheric Dilution (CAD) System }
TS 4.7.A.7.a Page 157b .
TS 4.7.A.7, item a, requires: The post-LOCA Containment Atmospheric !
Dilution (CAD) System be functionally tested once per operating cycle.
The post-LOCA CAD System provides the capability to inject nitrogen into :
the containment to ensure the oxygen-hydrogen mixture remains below the flammable limit following a postulated design basis LOCA. By design, ;
l the system is passive and normally maintained in the standby mode and contains suitable redundancy in components and features to ensure its l reliability. The H2 analyzers portion of the system are tested monthly ,
l to ensure the system's readiness. The valves used for venting and ;
nitrogen makeup are tested as part of our leak test program. . Little deterioration is experienced on the system due to its limited use.
Operability testing on the system was originally established on a l once/ cycle frequency based on the system redundancy. ;
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- Therefore, changing the CAD system functional test frequency from 18 months to 24 months will not affect the capability of the system to i
perform its design function. Our review of historical maintenance and surveillance data confirms this conclusion.
Standby Gas Treatment System (SGTS)
TS 4.7.B.l.a. (1), (2), (3), (4) Page 158 l TS 4.7.B.l.a requires: A demonstration of the following conditions at least once every 18 months: (1) Pressure drop across the combined HEPA filter and charcoal adsorber banks is less than 8 inches of water at ;
4000 cfm. (2) Inlet heaters are capable of providing at least 14KW )
output, (3) The HEPA filters have a 199% DOP removal rate, the charcoal adsorber banks have a 199% halogenated hydrocarbon removal rate, and a methyl iodide removal rate of 195% at specified environmental conditions, and (4) automatic initiation of each branch of SGTS.
The Standby Gas Treatment System is designed to filter and exhaust the reactor building atmosphere to the stack during secondary containment isolation conditions. The system is normally in the standby condition; thus, gross plugging or fouling of the HEPA filters and the charcoal adsorbers is minimized. The SGTS system has dual filter trains and heaters ensuring system availability in the event of failure of one of l
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l the system branches. The trains are tested monthly, to verify system !
operability and identify component degradation. A review of plant i surveillance records indicate that the measured parameters of the DP, j fl ow, charcoal efficiency, heater performance and HEPA performance do not vary significantly with time. The surveillance history of the !
automatic initiation test for each train of SGTS similarly justifies an ;
extension to 24 months. Based on the redundancy of the system design !
and review of past surveillance data, increasing the surveillance l interval to 24 months will not affect the capability of the system to j perform its design function. j i
Control Room Hiah Efficiency Air Filtration System (CRHEAF) ;
, TS 4.7.B.2.a Page 158B l l
TS 4.7.B.2.b Page 158B l TS 4.7.8.2.c Page 158C i TS 4.7.B.2.d Page 158C ,
The various surveillances of 4.7.B.2 currently require that once every 18 months certain tests be performed on the CRHEAF. These tests i demonstrate pressure drop across the HEPA filters and operability of the ,
filters and charcoal adsorbers in addition to testing the CRHEAF heaters ,
and humidistats.
l The CRHEAF is designed to filter intake air for the control room !
atmosphere during conditions when normal intake air may be contaminated. ;
The system is normally in standby condition; thus, gross plugging or ;
fouling of the HEPA filters and charcoal adsorbers will be minimized. !
In addition, the CRHEAF has redundant filter trains and fans that ensure !
system availability in the event of a failure of one of the system i components. Each train is tested monthly throughout the cycle to verify system operability and identify component degradation. A review of i
, plant surveillance records indicates that the measured parameters of Dp, j l fl ow, charcoal efficiency, heater performance, HEPA filter, and i humidistat performance do not vary significantly with time. Based on the redundancy of the system design and review of past surveillance ,
data, increasing the surveillance interval to 24 months will not affect '
the capability of the system to perform its design function.
! TS 4.7.C.I.c Page 159 l
TS 4.7.C.I.c requires: The Secondary containment capability to maintain 1/4 inch of water vacuum under calm wind (<5 mph) conditions with a filter train flow rate of not more than 4000 cfm be demonstrated each refueling outage prior to refueling.
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The secondary containment is designed to minimize any ground level release of radioactive materials which might result from serious accident. The intention of the specification is to assure the Secondary Containment is intact (or operable) prior to relying upon it for fuel handling. Performing these tests prior to refueling demonstrates secondary containment capability prior to the time the primary containment is opened for refueling. The frequency is event based Page 11-of 15
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j (i.e., prior to refueling) and not time dependent. The change to a 24
) month surveillance frequency will not alter the intention of the technical specification. Plant configuration, existing administrative controls, and PNPS procedures ensure Secondary Containment integrity is i
! maintained during normal operation. Periodic testing gives sufficient I 1 confidence of reactor building integrity and standby gas treatment system performance. Therefore, extending the fuel cycle will not L increase the risk of Secondary Containment degradation nor affect the
, capability of secondary containment to perform its design function. '
- Auxiliary Electrical System TS 4.9.A.1 Page 194 to 194A !
TS 4.9.A.2 Page 195 j Section 4.9.A.1 requires: Once per-operating cycle testing of various q Emergency Diesel Generator (EDG) functioning and loading. i
. The criteria for testing the Diesel Generators "once per operating l
- cycle" to show they will start, accept load, and, when tripped, will '
a automatically connect to offsite AC power is based on IEEE STD 338-1987.
IEEE STD 338-1987 states test intervals should be based on regulatory requirements, the scheduled plant operating cycle, and the testing's ;
impact on plant safety. ,
i i" A review of past testing for Pilgrim's EDGs shows no historical failures l for the performance at the current once-per-cycle 18 month testing i i interval. However, our calculation showed that extending' the testing 7 interval to 24 months would require a more accurate relay than currently installed to ensure the specified time sequence for starting and {
accepting the emergency load remains within specification. Therefore, we are changing two Diesel Generator breaker time delay relays whose function is to initiate EDG breaker closing. These relays will be replaced prior to extending the EDG test interval. Based on the ;
surveillance history and relay replacement, changing the test interval
] from 18 to 24 months will not adversely affect the EDGs and does not conflict with the epplicable EDG standard.
Section 4.9.A.2 requires: Once each operating cycle the batteries shall be subjected to a Service Discharge Test, and every 5 years the batteries must be subjected to a Performance Discharge Test (capacity).
- 4 Both the Service Discharge and Performance Discharge tests are derived from IEEE Std 450-1987. The Performance Discharge Test is on a five year cycle and is not impacted by the proposed change. The Service Discharge test has no frequency specified in IEEE Std 450-1987. .
Therefore, we reviewed increasing the test interval to 24 months using i the IEEE Std 338-1987 criteria as discussed above in T.S. 4.9.A.1 and determined that increasing the interval does not produce additional or accelerated battery degradation; the batteries are not prone to or known to sustain precipitous failures that a shorter interval may detect; and the most beneficial time to schedule the test (s) is during an outage.
i Therefore, increasing the interval to 24 months does not adversely affect the batteries and does not conflict with the applicable IEEE standards.
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Core Alterations TS 4.10 Pages 202 to 203a TS 4.10 requires: Verification of the operability of instrumentation and interlocks used in refueling and core alterations.
The requirements of section 4.10 are associated with core alterations during refueling. The requirements to surveil various protective instruments and interlocks prior to and during core alterations are event driven; that is, the interlocks and instruments are only required when the event of core alterations is imminent or in progress.
Therefore, increasing the interval to 24 months will not adversely impact the refueling and core alteration instrumentation and interlocks.
Alternate Shutdown Panels TS 4.12 Page 206 TS 4.12 requires: The demonstration of alternate shutdown panel operability once each cycle.
This test verifles the Alternate Shutdown Panels are capable of effecting safe shutdown in the event of a fire in the Cable Spreading Room. The nature of the hardware and components associated with the panels allowed the selection of the once/ cycle (18 month) testing l frequency, as failures were not expected to occur within that interval.
A review of surveillance history shows no time related failures associated with the existing 18 month interval. The basic components comprising the panels have proven themselves to be reliable and capable of performing as originally designed. No time-related failures were identified through a review of surveillance history. Thus extending to l a 24 month interval will not impact the availability and reliability of the alternate shutdown panels to function as designed.
D. Safety Evaluation and Determination of No Sionificant Hazard Considerations The Code of Federal Regulations (10CFR50.91) requires licensees requesting an amendment to provide an analysis, using the standards in 10CFR50.92, to determine whether a significant hazards consideration exists. The following analysis is provided in accordance with 10CFR50.91 and 10CFR50.92 for the proposed amendment. j
- l. The operation of Pilgrim Station in accordance with the proposed l amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated.
This submittal results in changes to various equipment surveillance intervals and related instrument calibration frequencies and setpoints.
The impact of lengthening the current 18 month interval to 24 months was evaluated and identified no significant system or component Page 13 of 15
I degradation as a consequence of lengthening the interval to 24 months; therefore, systems and components will continue to perform their design function. In some cases, the 24 month interval required setpoint changes to ensure instrument drift associated with the extended interval would not result in exceeding an instrument's acceptable setpoint tolerance. In other cases, justification of an extended interval was not developed because the surveillance could be performed on-line. In these cases, the "once/ cycle" surveillance requirement is changed to the currently allowed 18 months. !
The proposed changes were developed using the guidance provided in :
l Generic Letter 91-04 and Note 1 of Table 4,2. A through 4.2.G of !
l Pilgrim's technical specifications. The proposed changes do not :
i degrade the performance or increase the challenges to the associated ,
I safety systems assumed to function in the accident analyses. l i
The impact of lengthening the current calibration / functional test interval from 1 to 3 months for certain components was al so :
evaluated. The evaluation used the guidance in Generic Letter 91-04 :
and Note 1 of Table 4.2.A through 4.2.G of Pilgrim's technical !
specifications. No significant system or component degradation was ;
identified as a consequence of lengthening the interval to 3 months. l The proposed changes do not affect the availability of equipment or !'
systems required to mitigate the consequences of an accident, and do not affect the availability of redundant systems or equipment. The i plant will continue to operate within the limits specified in the ,
Core Operating Limits Report (COLR) and will continue to take the l same actions if setpoint limits are exceeded. "
F Therefore, both the proposed setpoint and non-setpoint changes do !
not significantly increase the probability or consequences of an !
accident previously evaluated. j i
- 2. The operation of Pilgrim Station in accordance with the proposed ~
amendment will not create the possibility of a new or different kind l of accident from any accident previously analyzed.
1 ,
The proposed changes with one exception, do not add or remove active !
components and, therefore, do not introduce failure mechanisms of a ;
different type than those previously evaluated. In one case, the !
EDG breaker time delay relays will be replaced with more accurate relays to ensure the specified time sequence for starting and ;
accepting the emergency load remains within specification for the ;
extended cycle. The replacement relays will be similar in size, i weight, voltage and temperature operating range as those being '
replaced; therefore, the possibility of a new or different kind of accident is not created. In addition, the surveillance test ;
requirements and the way surveillance tests are performed will remain unchanged. Since the intended operation and function of the i analyzed systems do not change as a result of the setpoint and non-setpoint analyses, no new initiators are introduced capable of 1 initiating an accident that would render these systems unable to provide their required protection. Therefore, the proposed changes l
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do not create the possibility of a new or different kind of accident :
from any accident previously evaluated. !
- 3. The operation of Pilgrim Station in accordance with the proposed ;
amendment will not involve a significant reduction in the margin of. ,
safety. !
i Although the proposed Technical Specification changes will result in an increase in the interval between surveillance tests, the existing :
margins of safety are maintained through our proposed setpoint !
revisions. The proposed setpoint changes either increase the plant j safety margin or maintain the existing margin and do not ;
significantly impact the availability, performance, or intended :
function of the affected systems. In the case of non-setpoint l changes, evaluation of the affected systems indicates lengthening the interval to 24 months does not have significant impact on !
performance. Therefore, the assumptions in Pilgrim's accident 4 analyses are not impacted, and the proposed Technical Specification changes do not significantly reduce the margin of safety.
This proposed change has been reviewed and recommended for approval by !
the Operations Review Committee and reviewed by the Nuclear Safety i Review and Audit Committee. l l
E. Schedule of Chance '
This is the third and final change reque*,t supporting the adoption of a 24 month refueling cycle at Pilgrim. We request the NRC review this proposed change in conjunction with the prior two and issue all the changes in one amendment. Implementation of the extended surveillance intervals will in some cases be put into affect within 60 days of receipt of approval of the amendment. For instrumentation requiring setpoint changes, implatentation of the extended surveillance intervals will not be put into ef fect until the changes are made. Attachment E identifies this instrumentation.
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