ML20207C227
| ML20207C227 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 02/22/1999 |
| From: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20207C235 | List: |
| References | |
| NUDOCS 9903080395 | |
| Download: ML20207C227 (19) | |
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_ Dave Morey
$1uthern Nucieer -
L Vice President :
' Operating Company farfey Project '
P.O. Box 1295 '
Birmingham Alabama 35201-Tel205 992.5131:
February 22, 1999 SOUTHERN comma Energy toServeYourWorld" Docket Nos.:. 348 NEle99-0058 -
50-364 l
U. S. Nuclear Regulatory Commission ATfN: ' Document Control Desk
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Washington, D. C. 20555-0001
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Joseph M. Farley Nuclear Plant 1
Request For Technical Specification Changes Control Room, Penetration Room, and Containment Purge Filtration Svetame and Fadiatian Manitorina Instrumentatim i
i Ladies and Gentlemen:
3 On June 30,1997, in accordance with the provisions of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) submitted Technical Specification (TS) amendments to clarify Surveillance Requirement (SR) references to ANSIN510 sections 10,12, and 13 to ASME N510-1989, ' Testing of Nuclear Air Treatment Systems," with errata dated January 1991, and to add a footnote which references the FNP Final Safety Analysis Report (FSAR) for relevant testing details. 'Ihe FNP FSAR is being revised to include a detailed discussion of the applicability of ASME N510-1989 sections 10,11, and 15. Differences between ANSI N510-1980 and ASME N510-1989 have been reviewed by SNC, and conversion to ASME N510-1989 for sections 10,11, and 15 is considered to be an enhancement. In addition, ASME N510-1989 is referenced by the
.J NRC in NUREG 1431.
j Additional related changes were proposed to reflect the latest safety analyses for a fuel handling accident (FHA) in the spent fuel pool (SFP) area or inside containment, and for post-LOCA Emergency Core Cooling System (ECCS) recirculation loop leakage outside containment. The Penetration Room Filtration System (PRFS) and Control Room Emergency Filtration System (CREFS) TSs were revised to provide " dirty filter" pressure and flow requirements and to revise
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charcoal adsorber laboratory testing efficiency criteria.
j The Containment Purge Emergency Filtration (CPEF) TS was deleted to reflect reliance upon rapid isolation of the containment by the purge exhaust radiation monitors. Radiation Monitoring.
J Instrumentation Table 3.3-6 was aaka d o require two channels of radiation monitors for the t
. containment purge, spent fuel pool area normal exhaust, and control room normal intake flow g
- Paths.
t 9903080395 990222 PDR ADOCK 05000348 l
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Page 2 U.S. Nuclear Regulatory Comminion
%is letter submits a complete replacement for the June 30,1997, submittal. Additional changes includui in this submittal are corrections to a typographical error in the typed pages previously provided, editorial changes to index pages, revision to the safety assessment to reflect the changes in the dose calculations resulting from responses to the Power Uprate Submittal RAls, changes resulting from questmns discussed in conference calls with the NRC Staff on October 5 & 13, 1998, wider flow rate acceptance bands, and reduction of the heater capacity in the CREFS.
Enclosure I provides a revised safety assessment for all the propW changes. Enclosure 2 provides a revision to the basis for a determination that the proposed changes do not involve a significant hazards consideration pursuant to 10 CFR 50.92. Enclosure 3 provides revised typed changes to the Unit 1 TSs. Enclosure 4 provides revised typed changes to the Unit 2 TSs. provides the Units 1 and 2 revised marked-up TS pages. De proposed Current Technical Specifications (CTS) change, including the supporting technical basis and significant hazards considerations, is also applicable to the proposed Farley Improved Technical Specifications (ITS) changes provided by SNC letter dated March 12,1998. Enclosure 6 provides marked-up typed pages of the Farley Improved Technical Specifications reflectmg any clanges required as part of this submittal. Review and approval of this licensing an-d==t change request is applicable to the ITS version as well, which will be incorporated into the ITS submittal when approved. Enclosure 7 contains information requested during conference calls with the NRC Staff condected on October 5 & 13,1998.
SNC has determined the proposed changes to the FNP TS do not involve a significant hazards consideration as defined in 10 CFR 50.92. De basis for this evaluation is provided in Enclosure i
- 2. SNC has also determined that the proposed changes will not significantly affect the quality of the human environment. A copy of the changed pages to insert into the proposed changes has been sent to Dr. D. E. Williams, the Alabama State Designee, in accordance with 10 CFR 50.91(b)(i).
J SNC requests that the NRC review and approve the proposed TS changes by July 1,1999. SNC plans to implement the proposed TS changes within 60 days ofissuance by the NRC.
Mr. D. N. Morey states that he is a Vice President of Southern Nuclear Operating Company and is authorized to execute this oath on behalf of Southern Nuclear Operatmg Company and that, to the best of his knowledge and belief, the facts set forth in this letter and enclosures are true.
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Page 3 U. S. Nuclear Regulatory Commission l
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If you havn any questions, please advisc.
d Respectfully submitted, SOUTIGRN NUCLEAR OPERATING COMPANY f/??h14l Dave Morey Sworn to and subscribed before me thistA0 of 999 N//dxf-Ia.
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kal b 1 lo NO/
My Commission Expires:
L JGS/maf: filtsrev. doc
Enclosures:
- l. Revised Safety Assessment
- 2. Revised 10 CFR 50.92 Evaluation
- 3. Unit 1 Revised Technical Specification Pages
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Unit 2 Revised Technical Specification Pages j
- 5. Units 1 & 2 Revised Marked-Up Technical Specification Pages l
- 6. Units 1 & 2 Typed and Marked-Up Farley Improved Technical Specifications Pages
- 7. Additiond Infonnation Requested From Conference Calls l
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Page 4 U. S. Nuclear Regulatory Commission cc:
Southern Nectar Operatian Comnany Mr. L. M. Stinson, General Manager - Farley U. S. Nucler Renu!9ary Commissiort Washinnton. D. C.
Mr. J. I. Zi.i i.crinan, Licensing Project Manager - Faricy U. S. Nucler Reentatary Commission. Renion II Mr.L. A.Reyes, Regional Administrator Mr. T. P. Johnson, Senior Resident Inspector - Farley Alabama Deoartmeni of Public Health Dr. D. E. Williamson, State Health Officer i
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N Joseph M. Farley Nuclear Plant Control Room, Penetration Room, and Containment Purge Filtration Systems and Radiation Monitoring Instrumentation Technical Specification Changes Revised Safety Assessment 1
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Joseph M. Farley Nuclear Plant Control Room, Penetration Room, and Containment Purge Filtratior Systems j-and Radiation Monitoring Instrumentation 1
Technical Specification Changes a
Revised Safety Assessment J
The Farley Nuclear Plant (FNP) Technical Specifications (TSs) for TS 3/4.7.7, Control Room Emergency l
Filtration System (CREFS), TS 3/4.7.8, Penetration Room Filtration System (PRFS), TS 3/4.9.12, j
Storage Pool Ventilation (Fuel Storage), TS 3/4.9.13, Storage Pool Ventilation (Fuel Movement), and TS 3/4.3.3, Radiation Monitoring Instrumentation are proposed to be revised TS 3/4.9.14, Contammer:
Purge Exhaust Filter (CPEF), is proposed to be deleted.
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FNP TS Surveillance Requirements (SRs) currently reference ANSI N510-1980 for performing in-place DOP testmg (section 10), charcoal adsorber leak testmg (section 12), and verifying laboratory testmg efficiencies (section 13) for FNP ventilation and filtration systems. Specific sections within ANSI N510-i 1980 do not clearly differentiate between testing required for ir/tial acceptance testmg and testmg required for periodic surveillances. In addition, some characteristics of tle FNP system designs do not allow for i
complete application of the 1980 standard without major modification or disassembly or significant breachmg of pressure boundaries. Testmg HEPA and charcoal adsorber combined pressure drop at design flow rate must be revised since the original system designs were not required to be in conformance with ANSI N509 as assumed by ANSI N510-1980. Adsorber efficiency laboratory testing in accordance with
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ASME NS 10-1989 recamm=AM methods does not require a large safety factor, thus the acceptance criteria are being revised 'Ibe CREFS Pressurization System filter heater output has been revised to l
eliminate excess capacity. The reduced heater capacity will provide ahd heater control functions while maintaining the necessary humidity control of the process air stream The safety analyses for post-LOCA ECCS recirculation loop leakage outside containment and fuel handling accident (FHA) in the sperA fuel pool area have been revised to be consistent with uprate analyses.
As an enhancement, a requirement to verify the capability of the PRF in the post-LOCA mode to maintain a negative pressure in the penetration room boundary is being added to the TSs.
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The revised safety analyses take credit for radiation monitoring instrumentation initiating protective actions in the event of a FHA. Two control room radiation monitoring channels will be required to provide redundant, single failure proofisolation of the normal HVAC system. For a FHA in containment, two channels of contamment purge and exhaust isolation radiation monitors will be required to provide redundant, single failure proofisolation of the containment with no credit for filtration, and for a FHA in the fuel storage pool area, two channels of fuel storage pool area radiation monitors will be required to provide isolation of the normal HVAC system and initiation of the PRF. These requirements for two i
channels are in response to a FHA, thus they are applicable only when moving irradiated fuel or moving heavy loads overirradistat fuel.
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l The following is a summary of the propom! TS changes:
- 1. TS SRs that currently reference ANSI N510-1980 sections 10,12, and 13 will be changed to make -
reference to ASME N510-1989. Wherever ASME N510-1989 is used in the TS SRs a footnote will be added that states.
"The FNP Final Safety Analysis Report identifies the relevant surveillance testmg requirements."
- 2. The pressure drop (measured in inches water gauge) across the combined HEPA and charcoal adsorber will be revised to a value consistent with the design of the system (2.3 inches for the CREFS Recirculation Unit,2.9 inches for the CREFS Filtration Unit,2.2 inches for the CREFS Pressurization Unit, and 2.6 inches for the PRF).
- 3. The adsorber laboratory testing criteria will be revised to be consistent with the testing methods rw = ='+i by ASME N510-1989, except that ASTM D3803-1989 will be used for laboratory testing 'of adsorber samples. The temperature for testing of the adsorber sample will be changed to 30*C for all filters; the efEciencies for the CREFS filters will be changed to 97.5%, 97.5% and 99.5% for the recirculation, filtration and pressunzation units respectively. A specific reference to ASTM D3803-1989 is being added to the surveillance requirement.
- 4. The flow rate for the CREFS and the PRF systems are being revised to increase operating flow ranges while ramalaing within the design filter system capacities.
- 5. The PRF heater dissipation surveillance will be deleted and operation time for the PRF will be revised from 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every 31 days on a staggered test basis to 15 minutes every 31 days on a staggered test basis.
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- 6. As a clarification, the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> run with the heater control circuit energized on the control room pressurization unit has been changed to be consistent with the Improved Technical Specification (ITS) l wording.
- 7. As an aahancamant, an additional surveillance to venfy the integrity of the PRF boundary will be added.
j New surveillance requirement 4.7.8.e will &=--wm that one PRF system remains capable of maintaining the RHR heat exchanger room at a negative pressure s -0.125 in, water gauge, and the remaining surveillances are renumbered appropriately.
- 8. Footnotes will be added to Table 3.3-6 notmg that two chana*Ie of control room radiation monitors, containment purge and exhaust radiation momtors, and fuel storage pool area radiation monitors are required when moving irradiated fuel or heavy loads over irradiated fuel. Action 25 will be revised consistent with the flexibility provided by the existmg PRF =ctn= tai signal design.
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- 9. Action 27 of Table 3.3-6 will be clarified to require operation of the control room pressurization as well as the recirculation subsystems in the emergency recirculation mode.
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- 10. De containment purge exhaust Ster will be removed from the TSs based on new dose analyses.
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- 11. A clarine=*iaa is being added to the action statements for 3.9.12 and 3.9.13 to clarify that the actions apply to the movement ofirradiated fuel and not new fuel.
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- 12. The bases for affected TSs will be revised consistent with the above changes, and system names will i
be revised consistent with FNP nomenclature. Dere are also two editorial changes which remove I
footnotes tlut are no longer applicable.
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- 13. Excess beater capacity will be deleted from the CREFS Pressurization System filter heaters to prevent j
heater cycling on the thermal cutouts during the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> surveillance test.
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Basis for the proposed changes above:
i Change 1 - It is the intent of the first proposed change to clarify the requirements associated with FNP j
filtration system testing. Conversion of the ANSI N510-1980 sections 10,12, and 13 to ASME N510-j 1989 sections 10,11, and 15 will bring FNP SRs closer to current industry standards. The relae=*iaa of j
specific testmg requirements to the FSAR is consistent with guidance provided by NUREG 1431, Improved l
Standard Technical Specifications for Weinghe Plants. FNP has submitted an ITS package by SNC l
letter dated March 12,1998, and clean typed and marked up pages to that submittal are included as A*=ch=-a* 6.
i-Differences between ANSI N510-1580 and ASME N510-1989 have been reviewed by SNC, and j
conversion to ASME N510-1989 seconn= 10,11, and 15 for ANSI N510-1980 sections 10,12, and 13 are considered to be an enhancement. In addition, ASME N510-1989 is..&. ad by the NRC in NUREG l
1431. Inclusion of the specific testing requirements in the FNP FSAR will ensure that any deviation to the testing requirements of ASME N510-1989 will receive appropriate review through the 10 CFR 50.59 i
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process for this change and any changes made in the future.
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Change 2 - Proposed change 2 is required because some characteristics of the FNP system designs do not -
allow for complete application of the 1980 standard without major modifications Testing HEPA and i
charcoal adsorber combined pressure drop at design flow rate must be revised to =A~==* ly verify " dirty" filter pressure drop limitations since the original fan and system designs were required to be in conformance
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with Regulatory Guide 1.52 in lieu of ANSI N509 as assumed by ANSI N510-1980. He revised values reflect the as-installed dirty filter pressure drop limitations of the FNP equipment. Verification of these 3
values at design flow will maintain the filter within design internal pressure loads.
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Change 3 - Changing the filter test whadalogy to be consistent with ASME N510-1989 includes a comnutment to ASTM D3803-1989. His revision of the standard is-- =4= a charcoal test ts.,widare of 30 *C which is consistent with the FNP filter operatmg tempemture and will be =AapM In support of proposed change 3, revised safety analyses have been prepared for the ECCS recirculation loop leakage outside containment contributions to offsite and convoi room LOCA doses and offsite doses for fuel handhng accidents in the fuel storage poo! area and in the e<=t=inment.
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Using the data shown in Table 1 yields total doses of 26.9,180, and 105 REM thyroid for the control l
room, site boundary and Low Population Zone (LPZ) respectively, which continue to meet General Design Criterion (GDC) 19 and 10 CFR 100 guidelines. Using the assumptions shown in Table 2, the resultant '
FHA doses are shown in Table 3. %ese results meet the Standard Review Plan (SRP) Section 15,7.4 j
criteria of maintaining offsite doscs well within 10 CFR 100 guidelines.
l Recent discussions with the NRC staff have ladi* that the safety factors implied in Regulatory Guide j
1.52 are overly conservative when applied to the conservative test methodology is = '+i by ASME L
i N510-1989 (and by reference, ASTM D3803-1989) and a safety factor of two would be acceptable.
j Hence, the laboratory test acceptance criteria for CREFS efficiencies credited in the safety analyses (94.5%
l for the 2 inch deep recirculation and Stration units and 98.5% for the 6 inch deep pressunzation unit, j.
including 0.5% reduction for bypass leakage) are revised to refic< a safety factor of two. His yields laboratory test acceptance criteria of 97.5% for the recirculation and filtration units and 99.5% for the j
pressurization unit. This safety factor and the conservative test methods and dose calculations ensure that j.
control room operator doses will continue to meet GDC 19 limits. Also, a safety factor of 2 is being used j
for the PRF which yields a laboratory test acceptance criteria of 95%.
f Change 4 - In order to provide operating flexibility of the CREFS and the PRF filter systems, change 4 makes the flow rate band wider. De CREFS Pressurization flow is being increased from 300 CFM i 10%
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to 375 to 270 CFM. He increased flow rate tolerance is well within the design capability of the system j
and assures the system face velocity will be less than the lab testmg velocity. Dose calculations conservatively use the maximum flow rate of 125% of nominal design flow. De lower bound is the value j
used to establish the dirty filter differential pressure. PRF flow is being increased from 5000 CFM 10 %
to 6000 to 4500 CFM. He PRF upper bound is within the design capacity of the system and the kmer bound is the value used to establish the dirty filter differential pressure.
l Change 5 - Since the revised safety analyses indicate that dose results meet acceptance criteria without credit for the PRF beater, change 5 proposes to delete the heater dissipation surveillance, and delete the 10 l
hour heater run time and replace it with a verification that the system performance is stable for 15 minutes.
l This testing requirement will venfy the flow stability and alignment of the system and is consistent with guidance provided by NUREG 1431, Improved Stmadard Technical Speci5 cations for W~tiagha-j Plants, for Siter systems without heaters. %c PRF system is also run for 15 minutes in the fuel handling j
accident alignment per TS SR 4.9.12.2 a.
Change 6 - To clarify the wording on the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> run for the control room pressurization unit, change 6 j
adds the wording from NUREG 1431 on the heater control circuit being energized during the run. His change will bring this surveillance into agreement with the ITS.
Change 7.- As an aah= ament of the testing program, and in response to concerns about verification of the condition of the imdion room pressure boundary integrity, change 7 will add a surveillance of the capability of each PRFS to maintain a negative pressure in the RHR heat ~ hanger room with respect to adjacent areas. His requirement will provide rea=anahle assurance that the penetration room pressure boundary has not suffend degradation, =taimiting unfiltered release of ECCS recirculation loop El-4 4
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4 leakage, and thereby providing additional confidence that offsite and control room doses will remam within i
10 CFR 100 and GDC 19 guidelines.
Change 8 ~Ihis change is proposed to conform the TSs with the revised accident analyses prepared to support changes 3,4, and 5 described above. The revised analyses take credit for operation of the affected radiation monitors; and in order to mitigate a single failure of one radiation monitor, a second radiation monitor must be available. Since the radiation monitors are modeled to mitigate the consequences of an FHA, two channels must be available when the possibility of an FHA exists; that is, when irradiated fuel is L being moved or when heavy loads are being moved over irradiated fuel No change to safety analyses which credit a radiation monitor was made for any other condition, so the requirement for two channels is limited to the conditions described here. As discussed in items 3,4, and 5 above, the offsite dose results continue to be well within the 10 CFR 100 guidelmes. Since either fuel storage pool area radiation monitor will provide isolation of the normal HVAC system and thereby geomte both trams oflow differential pressure signals on loss of the normal HVAC flow, both trains of PRF will receive start signals from proper functioning of either redsstinn monitor. This existing redundarry allows unhnking action 25 for the fuel storage pool at:a radiatinn monitor from the operability of the PRF filters; and actions simdar to the fuel storage pool ventilation actions are added here to provide the same level of protection as the current -
TSs.
Change 9 - Action 27 to Table 3.3-6 is being revised per change 9 in order to clarify that protection of the control room requires that the recirculation, filtration, and the pressurization filter units be placed in operation. This configuration, with all filters in one train operating, is consistent with the flow paths, flow rates, and filter functions modeled in the safety analyses. This is an editorial change to conform the TSs ternunology to that used in FNP procedures, which reduces the potennal for misuais Sag and thereby a
increases confidence in proper operation of the CREFS.
b Change 10 - To support proposed change 10, the FHA inside containrnent was re-analyzed with no credit for the containment purge and exhaust filter. In this case, the purge and exhaust radiation monitor will detect the activity released to the mntainment and isolate the purge sprem. The time required to detect and isolate the purge, shown in Table 2, includes the time to purge the actisity in the purge and exhaust ductwork downstream of the isolation valves. Other major parameters used in the analysis are also shown j
in Table 2. The results shown in Table 3 meet the SRP Section 15.7.4 criteria of maintaimng offsite doses i
well within 10 CFR 100 guidelines, without credit for filtration of the ^"M activity. Therefore, it is acceptable to delete the containment purge exhaust filter from the TSs.
Change 11 - Proposed change 11 is to clarify that the action statement applys to the movement ofirradiated fuel and not to new fuel. New fuel does not constitute a heavy load as 1 if a new fuel assembly is dropped there is no release of radioactivity. The use of the term irradiated fuel is also consistent with NUREG 1431.
Change 12 - Proposed change 12 is required to maintain consistency between the affected TSs and their Bases.
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i Change 13 - Proposed change 13 is required to enhance the heater controls and clinunate heater cyc!ing during surveillance tests of the CREFS Pressurization System filter heaters. Adequate heater capacity will be provided by the 2.5 kW capacity at the new maximum flow rate of 375 CFM to maintain the relative humidity of the inlet air to s 70 %.
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TABLE I (Sheet 1) l EVALUATION OF LEAKAGE FROM THE RECIRCULATION LOOP 4
Recirculation Loop h91992.
Concentration (uCi/nm) 3 i
I-131 2.7 x 10' I-132 4.0 x 10' I-133 5.9 x 10*
I134 6.2 x 10' I135 5.5 x 10' i
Parameters
. Values Power level (MWt) 2831 l
Equivalent percent fuel failure 100 Fraction ofiodine activity absorbed by 0.5 sump water 1
Sump water volume RCS (ft3) 9,107 RWST(ft3) 40,100 Leak Rate 10 x FSARTable 6.3-8 Fraction which flashes 0.1 PRF e&iency (')
Elemental / Organic Iodine 89.5 %
Particulate Iodine 98.5 %
(*) Includes 0.5% reduction for bypass leakage.
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i TABLE 1 (Sheet 2)
EVALUATION OF LEAKAGE FROM THE RECIRCULATION LOOP Control Room Parameters Volume 114,000 ft' Pressurization Flow 375 cfm Recirculation Flow 2,700 cfm Unfiltered Inleakage 10 cfm Pressurization Filter Efficiency 98.5 M')
Recirculation Filter EfBeiency 94.5 W')
es aspheric Dilution Factors 3.28 x 10 sl.n' 2.65 2.19 1.64 1.08 b
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4 kludes 0.5 % reduction for byWs IQ El-8
TABLE 2 PARAMETERS USED IN FUEL HANDLING ACCIDENT ANALYSIS -
Accidentin Accident in Fuel Storage Contamment Pool Area (Auxiliary Buildinsd Core thermal power (MWt) 2831 2831 Time from shutdown to accident (h) 100 100 Minimum water depth (ft) 23 23 Damage to fuel assembly All rods ruptured All rods ruptured Activity release from assembly Kr-85 30%
30 %
Other Noble Gases 10 %
10 %
I-131 12 %
12 %
Other Iodmes 10 %
10 %
Radial peaking factor 1.7 1.7 Decontamination factor in water Iodine -Elemental 133 133
- Organic 1
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Noble gases 1
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5 Amount of moung in building (ft')
6.6 x 10 1 x 10 Exhaust flow rate (cfm) 5.35 x 10 12000 d
Isolation time (sec) 45 None Iodine filtration system None.
PRF i
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Filter efficiency N/A 89.5 % ')
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j Atmospheric dilution factors (see FSAR (see FSAR 1
table 15B-2) table 15B-2) i l.
(*) - Includes 0.5 % reduction for bypass leakage El-9 J
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TABLE 3 i
OFFSITE DOSES FROM FUEL HANDLING ACCIDENT Accident in Accidentin Fuel Containmetit Storane Pool Area l
Site Boundary Dose (REM)
Thyroid 12.1 21.6 h ieBody 0.4 0.4 4
i Low-Population Zone Dose (REK, I
Thyroid 4.5 7.9 4
hie Body 0.1 0.1 i
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l Jaseph M. Farley Nuclear Plant Control Room, Penetration Room, and Containment Purge Filtration Systems and Radiation Monitoring Instrumentation Technical Specification Changes Revised 10 CFR 50.92 Evaluation t
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Joseph M. Farley Nuclear Plant Control Room, Penetration Room, and Containment Purge Filtration Systems and Radiation Monitoring Instrumentation Technical Specification Changes Revised 10 CFR 50.92 Evaluation f
Pursuant to 10 CFR 50.92, SNC has evaluated the proposed amendments and has determmed that operation of the facility in accordance with the proposed amendments would not involve a significant hazards considemtion. The basis for this determmation is as follows:
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- 1. De proposed changes do not involve a significant increase in the probability or consequences of an l
accident previously evaluated.
The proposed changes to convert from ANSI N510-1980 to ASME N510-1989 for specific FNP.
filtration surveillance testmg requirements and related changes do not affect the probability of any accident occurring. He consequences of any accident will not be affected since the proposed changes will continue to ensure that appropriate and required survedlance testing for FNP filtration systems will i
be performed consistent with the revised accident analyses. nc results of the fuel handling accident remain well within the guidelines of 10 CFR Part 100 and the doses due to a LOCA, including ECCS recirculation loop leakage, remam within the guidelines of 10 CFR Part 100 and General Design Criterion 19 of Appendix A to 10 CFR Part 50. Relocatmg specific testmg reqmrements to the FNP FSAR has no effect on the probability or consequences of any accident previously evaluated since required testmg will continue to be performed.
i Therefore, the proposed TS changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
Testing differences between ANSI N510-1980 and ASME N510-1989 have been evaluated by SNC and none of the proposed changes have the potential to create an accident at FNP. ASME N510-1989 is referenced by the NRC in NUREG 1431. Testing the addi ional channele of radiation monitoring t
and verification of penion room boundary integrity do not require the affected systems to be placed in configurations different from design. Bus, no new system design or testmg con 6guration is required for the changes being proposed that could create the possibility of any new or different kind of accident from any accident previously evaluated. Relocating specific testmg requirements to the FSAR has no effect on the possibility of creatmg a new or different kind of accident from any accident previously evaluated since it is an admi%tive change in nature.
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t 10CFR50.92 Evaluation Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. The proposed changes do not involve a significant reduction in a margin of safety.
Conversion from the testing requirements of ANSI N510-1980 sections 10,12, and 13 to ASME NS 10-1989 sections 10,11, and 15 has been previously approved by the NRC at other nuclear facilities. ASME N510-1989 has been approved and endorsed by the NRC in NUREG 1431. The safety factor associated with the conservative charcoal adsorber laboratory test methods and dose calculations ensures that doses will contmue to meet the guidelines of 10 CFR Part 100 and GDC 19 of -
Appendix A to 10 CFR Part 50. The enhanced testmg of radiation monitoring instrumentation and the
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penetration room boundary integrity provide additional assurance that the acceptance criteria of the safety analyses and the resultant margins of safety are not reduced. Relocatmg specific testing requirements to the FSAR has no effect on the margin of plant safety since required testing will continue to be performed. Clarifymg the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> run with heaters on is consistent with the Improved TS language and accomplishes the purpose for the surveillance. Changing the heater capacity and flow rates has been factored into the dose calculations and are within the design capacities of the systems involved. Therefore, SNC concludes based on the above, that the proposed changes do not result in a significant reduction of margin with respect to plar afety as defined in the Final Safety Analysis Report or the bases of the FNP technical specificaLms.
'Iberefore, the proposed changes do not involve a significant reduction in a margin of safety.
Conclusion i
Based on the preceding analysis, SNC has determined that the proposed changes to the Technical Specifications will not significantly increase the probability or consequences of an accident predously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety. SNC therefore concludes that the proposed changes meet the requirements of 10 CFR 50.92(c) and do not involve a significant hazards consideration.
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Joseph M. Farley Nuclear Plant Control Room, Penetration Room, and Containment Purge Filtration Systems and Radiation Monitoring Instrumentation Technical Specification Changes Unit 1 Revised Technical Specification Pages
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