L-99-033, Application for Amends to Licenses NPF-2 & NPF-8,to Change TS Re Control Room,Penetration Room & Containment Purge Filtration Sys & Radiation Monitoring Instrumentation,Which Reflect Agreements Reached in 990909 & 16 Discussions
| ML20217B157 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/04/1999 |
| From: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20217B161 | List: |
| References | |
| NEL-99-0336, NUDOCS 9910120167 | |
| Download: ML20217B157 (17) | |
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D:ve Morey Southern Nuclear Vice President - Operating Company Farley Project P0 Box 1295 -
Birmingham, Alabama 35201 Tel 201992.5131 October 4, 1999 EM L COMPANY Energy to Serve YourWorld" Docket Nos.: 50-348 NEL-99-0336 50-364 i
U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C.20555-0001 :
i Joseph M. Farley Nuclear Plant Request For Technical Specification Changes Control Room, Penetration Room, and Containment Purge filtration Systems and Radiation Monitorina Instrumentation Ladies and Gentlemen:
By letters dated June 30,1997, February 22,1999, March 19,1999, and June 30,1999, in accordance with the provisions of 10 CFR 50.90, Southern Nuclear Operating Company (SNC) submitted Technical Specification amendments to revise Surveillance Requirement (SR) references from ANSI N510-1980 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. The 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 NRC in NUREO 1431.
This letter submits revised pages to the Technical Specification amendment that reflect agreements reached in discussions with the staff on September 9,1999 and September 16,1999. The changes provided in this letter are to keep the PRF flow tolerance band to i 10 %, add a negative pressure i sutveillance on the Spent Fuel Room, and to add clarification on the Emergency Recirculation Mode of opration plus a description on heater controls to the Technical Specification bases for the Control Room Emergency Filtration System (CREFS) Pressurization Unit.
Enclosure i provides a revised safety assessment for all the proposed changes which has the changes 3g\
marked v,ith a side bar. Enclosure 2 provides a revision to the basis for a determination that the 7' 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 Technical Specifications. Enclosure 4 provides revised typed changes to the Unit 2 Technical Specifications. Enclosure 5 prosides the s Units I and 2 revised marked-up Technical Specification pages. The proposed Current Technical 9910120167 991004-PDR ADOCK 05000348 P PDR-
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. Specifications (CTS) change, including the supporting technical basis and significant hazards consideratiws, is also applicable to the proposed Farley improved Technical Speci6 cations (ITS) changes prorvided by SNC letter dated March 12,1998. Enclosure 6 provides marked-up and typed pages of the Farley improved Technical Specifications reficcting any changes required as part of this submittal. Review and approval of this licensing amendment change request is applicable to the ITS version as well, which will be incorporated into the ITS submittal when approved. Enclosure 7 provides SNC responses to the Request For Additional Information letter dated September 30,1999.
SNC has determined the proposed changes being submitted with this letter do not involve a j significant hazards consideration as defined in 10 CFR 50.92. The basis for this evaluation is l provided in the revised Enclosure 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 insent into the proposed changes has been sent to Dr. D. E. Williamson, the Alabama State Designee, in - .
accordance with 10 CFR 50.91(b)(i). J Mr. D. N. Morey states that he is a Vice President of Southern Nuclear Operating Company and is l authorized te execute this oath on behalf of Southern Nuclear Operating Company and that, to the l best of his knowledge and belief, the facts set forth in this letter and enclosures are true.
If you have any questions, please advise.
Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY fl )NfN Dave Morey Sworn to andsubscribedbefor me this day of 1999 7a/axthu
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Enclosures:
i 1, Revised Safety Assessment )
- 2. Revised 10 CFR 50.92 Evaluation ;
- 3. Unit 1 Revised Technical Specification Pages l "4. Unit 2 Revised Technical Specification Pages j
- 5. Units 1 & 2 Revised Marked-Up Technical Spc .ification Pages l
- 6. Units 1 & 2 Typed and Marked-Up Farley improved Technical Specifications Pages l
, 7. Response to RAI Questions i l
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.[ U. S. Nuclear Regulatory Commission cc: Southern Nuclear Ooeratina Comoany Mr. L. M. Stinson, General Manager - Farley U. S. Nuclear Renulatorv Commiss'on. Washinnton. D. C.
, Mr. L. M. Padovan, Licensing Project Manager - Farley U. S. Nuclear Renulatory Commission. Renion II Mr. L. A. Reyes, Regional Administrator Mr. T. P. Johnson, Senior Resident inspector - Farley Alabama Department of Public Health .
Dr. D. E. Williamson, State Health Officer i
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4 Enclosure 1 Joseph M. Farley Nuclear Plant Control Room, Penetration Room, and Containment Purge Filtration Systems and Radiation Monitoring Instrumentation Technical Specification Changes Revised Safety Assessment 8
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Enclos'ure 1
' Joseph M. Farley Nuclear Plant Control Room, Penetration Room, and Containment Purge Filtration Systems -
and Radiation Monitoring Instrumentation
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Technical Specification Chrmges
- Revised Safety' Assessment The Farley Nuclear Plant (FNP) Technical Specifications for Technical Specification 3/4.7.7, Control
' Room Emergency Filtration System (CREFS), Technical Specification 3/4.7.8, Penetration Room Filtration System (PRFS), Technical Specification 3/4.9.12, Storage Pool Ventilation (Fuel Storage),
sTechnical Specification 3/4.9.13, Storage Pool Ventilation (Fuel Movement), and Technical Specification .
3/4.3.3, Radiation Monitoring Instrumentation are proposed to be revised. Technical Specification 3/4.9.14, Containment Purge Exhaust Filter (CPEF), is proposed to be deleted.
FNP Technical Specification Surveillance Requirements (SRs) currently reference ANSI N510-1980 for performing in-place DOP testing (section 10), charcoal adsorber leak testing (section 12), and verifying laboratomy testing efficiencies (section 13) for FNP ventilation and filtration systems. Specific sections L within ANSI N510-1980 do not clearly differentiate between testing required for initial acceptance testing and testing required for periodic surveillances. In addition, some characteristics of the FNP system designs do not allow for complete application of the 1980 standard without major modification or disassembly or significant breaching of pressure boundaries. Testing 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 ASME N510-1989 recommended methods does not require a large safety factor, thus the acceptance criteria are being revised. The CREFS Pressurir.ation System filter heater output has been revised to eliminate excess capacity. The reduced heater capacity will provide enhanced 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 spent fuel pool area have been revised to be consistent with uprate analyses.
As an enhancement, requirements to verify the capability of the PRF in the post-LOCA mode to maintain a negative pressure in the penetration room boundary and to maintain a negative pressure in the spent fuel room in the fuel handling accident mode are being added to the Technical Specifications.
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 containment 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 channels are in response to a FHA, thus they are applicable only when moving irradiated fuel or moving
' heavy loads over irradiated fuel.-
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Enclosure 1 l
The following is a summary of the proposed Technical Specification changes:
- l. Technical Specification 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 Technical Specification SRs a footnote will be added that states: )
I "The FNP Final Safety Analysis Report identifies the relevant surveillance testing 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 adsoiber laboratory testing criteria will be revised to be consistent with the testing methods recommended by ASME N510-1989, except that ASTM D3803-1989 will be used for laboratory testing of adsorber samples. The temperatare for testing of the adsorber sample will be changed to 30 C for all filters; the efficiencies for the CREFS filters will be changed to 97.5%, 97.5% and 99.5% for the recirculation, filtration and pressurization units respectively. A specific reference to ASTM D3803-1989 is being added to the surveillance requirement.
- 4. The flow rate for the CREFS system is being revised to increase operating flow range while remaining l ;
within the design filter system capacities, j
- 5. The PRF heater dissipation surveillance will be deleted and operation time for the PRF will be resised 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.
- 6. As a clarirication, the bases has been changed to add wording on the CREFS pressurization unit heater operation. {
7 As an enhancement, an additional surveillance to verify the integrity of the PRF boundary will be added.
New surveillance requirement 4.7.8.e will demonstrate 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. l
- 8. Footnotes will be added to Table 3.3-6 noting that two channels of control room radiation monitors, containment purge and exhaust radiation monitors, 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 existing PRF actuation signal design.
- 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. The containment purge exhaust filter will be removed from the Technical Specifications based on new I dose analyses.
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- 11. A clarification 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. I
- 12. The bases for affected Tecimical Specifications will be revised consistent with the above changes, and .
system names will be revised consistent with FNP nomenclature. The bases have also been revised to !
add clarification on the emergency recirculation mode of operation for the control room filtration i systems and a description of the pressurization system heater operation. 'Ihere are also two editorial changes which remove footnotes that are no longer applicable and some grammar corrections.
- 13. Excess heater capacity will be deleted from the CREFS Pressurization System filter heatera to prevent 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.
- 14. A new surveillance is being added to the PRF system to verify negative pressure in the spent fuel room.
l Basis for the proposed changes above:
Change 1 - It is the intent of the first proposed change to clarify the requirements associated with FNP filtration system testing. Conversion of the ANSI N510-1980 sections 10,12, and 13 to ASME N510-1989 sections 10,11, and 15 will bring FNP SRs closer to current industry standards. The relocation of j specific testing requirements to the FSAR is consistent with guidance provided by NUREG 1431, Rev.1, !
Standard Technical Specifications Westinghouse Plants. FNP submitted an ITS package by SNC letter dated March 12,1998, and clean typed and marked up pages to that submittal are included as Enclosure 6.
1 Differences between ANSI N510-1980 and ASME N510-1989 have been reviewed by SNC, and I conversion to ASME N510-1989 sections 10, i1, and 15 for ANSI N510-1980 sections 10,12, and 13 is )
considered to be an enhancement. In addition, ASME N510-1989 is referenced by the NRC in NUREG 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 process for this change and any enanges made in the future.
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. Testim iliiPA and charcoal adsorber combined pressure drop at design flow rate must be revised to adequ acly verify " dirty" filter pressure drop limitations since the eriginal fan and system designs were required to be in conformance with Regulatory Guide 1.52 in lieu of ANSI N509 as assumed by ANSI N510-1980. The revised values reflect the as-installed dirty filter pressure drop limitations of the FNP equipraent. Verification of these values at design flow will maintain the filter within design internal pressure loads.
. Change 3 - Changing the filter test methodology to be consistent with ASME N510-1989 includes a commitment to ASTM D3803-1989. This revision of the standard recommends a charcoal test temperature of 30 *C which is consistent with the FNP filter operating temperature and will be adopted.
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Enclosure 1 In support of proposed change 3, revised safety analyses have been prepared for the ECCS recirculation loop leaka8e outside containment contributions to offsite and control room LOCA doses ar.d offsite doses for fuel handling accidents in the fuel storage pool area and in the containment. Using the data shown in Table 1 yields total doses of 26.9,180, and 105 REM thyroid ibr the control 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. These results meet the Standard Review Plan (SRP) Section 15.7.4 criteria of maintaining offsite doses well within 10 CFR 100 guidelines.
Recent discussions with the NRC staff have indicated that the safety factors implied in Regulatory Guide 1.52 are overly conservative when applied to the corervative test methodology recommended by ASME N510-1989 (and by reference, ASTM D3803-1989) and a safety factor of two would be acceptable.
Hence, the laboratory test acceptance criteria for CREFS efficiencies credited in the safety analyses (94.5%
for the' 2 inch deep recirculation and filtration units and 98.5% for the 6 inch deep pressurization unit, including 0.5% reduction for bypass leakage) are revised to reflect a safety factor of two. His yicids laboratory test acceptance criteria of 97.5% for the recirculation and filtration units and 99.5% for the pressurization unit. This safety factor and the conservative test methods and dose calculations ensure that control room operator doses will continue to meet GDC 19 limits. Also, a safety factor of 2 is being used for the PRF which yields a laboratory test acceptance criteria of 95%.
Change 4 -In order to provide operating flexibility of the CREFS filter system, change 4 makes the flow rate band wider. The CREFS Pressurization flow is being increased from 300 CFM 10% to 375 to 270
- CFM. The increased flow rate tolerance is well within the design capability of the system and assures the system face velocity will be less than the lab testing velocity. Dose calculations conservatively use the 1 maximum flow rate of 125% of nominal design flow. The lower bound is the value used to establish the l dirty filter differential pressure. . l Change 5 - Since the revised safety analyses indicate that dose results meet acceptance criteria w"hout ;
credit for the PRF heater, change 5 proposes to delete the heater dissipation surveillance, and delete the 10 1 I
hour heater run time and replace it with a verification that the system performance is stable for 15 minutes.
This testing requirement will verify the flow stability and alignment of the system and is consistcnt with guidance provided by NUREG 1431, Rev.1 Standard Technical Specifications - Westinghouse Plants, for l .
filter systems without heaters. The PRF system is also run for 15 minutes in the fuel handling accident I alignment per Technical Specification SR 4.9.12.2 a.
Change 6 - A description to clarify the operation of the heaters on we CREFS Pressurization System
' during the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> run is added to the bases sections of CTS and ITS. This clarification indicates that the heaters are on during the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> run unless the protective thermal cutouts operate to protect the charcoal from overheating.
Change 7 - As an enhancement of the testing program, and in response to concerns about verification of the condition of the penetration room pressure boundary integrity, change 7 will add a surveillance of the
- l. capability of each PRFS to maintain a negative pressure in the RHR heat exchanger room with respect to ;
i adjacent areas. This requirement will provide reasonable assurance tLat the penetration room pressure boundary has not suffered degradation, minimizing unfiltered release cf ECCS tecirculation loop El-4 i
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Enclosure 1 leakage, and thereby providing additional confidence that offsite and control room doses will remain within 10 CFR 100 and GDC 19 guidelines.
Change 8 - This change is proposed to conform the Technical Specifications 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 FIIA exists; that is, when irradiated fuel is 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 guidelines. Since either fuel storage pool area radiation monitor will provide isolation of the normal HVAC system and thereby generate both trains oflow differential pressure signals on loss of the normal HVAC flow, both trains of PRF will receive start signals from proper functioning of either radiation monitor. This existing redundancy allows unlinking action 25 for the fuel storage pool area radiation monitor from the operability of the PRF filters; and actions similar to the fuel storage pool ventilation actions are added here to provide the same level of protection as the current Technical Specifications.
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 patlu, flow rates, and filter functions modeled in the safety analyses. This is an editorial change to conform the Technical Specifications terminology to that used in FNP procedures, which reduces the potential for I
misunderstanding and thereby increases confidence in proper operation of the CREFS.
Change 10 - To support proposed change 10, the FHA inside containment was re-analyzed with no credit i for the containment purge and exhaust filter. In this case, the purge and exhaust radiation monitor will detect the activity released to the containment and isolate the purge system. The time required to detect and isolate the purge, shown in Table 2, includes the time to purge the activity in the purge and exhaust ductwork downstream of the isolation valves. Other major parameters used in the analysis are also shown in Table 2. The results shown in Table 3 meet the SRP Section 15.7.4 criteria of maintaining offsite doscs well within 10 CFR 100 guidelines, without credit for filtration of the exhausted activity. Therefore, it is acceptable to delete the containment purge exhaust filter from the Technical Specifications.
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 and 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 Technical Specifications and their Bases and to incorporate information requested by the NRC Staff. !
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l Chenge 13 - Proposed change 13 is required to enhance the heater controls and climinate heater cycling during surveillance test 3 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 %
Change 14 - Proposed change 14 is adding a negative pressure surveillance test to the PRF system Technical Specification to test the capability of the system to produce a slightly negative pressure in the spent fuel rocm when the system is operating in the fuel handling accident mode, This surveillance is similar to the change described in Change 7 above. It will demonstrate that the PRF system remains capable of maintaining the spent fuel room at a slightly negative pressure to assure that there is no unfiltered leakage.
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- TABLE 1 (Sheet 1) -
EVALUATION OF LEAKAGE FROM THE RECIRCULATION LOOP l
l Recirculation Loop Isotope Concentration (uCi/2m)
I131- 2.7 x 10' I-132' 4.0 x 10' '
I-133 5.9 x 10' L
l I-134 6.2 x 10' I-135 5,5 x 10' l
Ji L Pammeters Values
- Power level (MWt) 2831 Equivalent percent fuel failure - 100
. Fraction ofiodine activity absorbed by 0.5 sump water Sump water volume RCS (ft') 9,107 RWST (ft') - 40,100 ,
Leak Rate ' 10 x FSAR Table 6.3-8 ;
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Fraction which flashes 0.1 FRF efficiency W'-
Elemental / Organic Iodine'. 89.5 %
Particplate Iodine . 98.5 % l M t-ins. 0.5"Areduction for bypass Jwage.
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EVALUATION OF LEAKAGE FROM TiiE RECIRCULATION LOOP
' Control Room Parameters 2
Volume - 114,000 f1
! Pressurization Flow 375 cfm Recirculation Flow. 2,700 cfm Unfiltered Inleakage 10 cfin Pressurization Filter Efficiency 98.5 M')
Recirculation Filter EfYiciency 94.5 M')
Atmospheric Dilution Factors. 3.28 x 10' s/m' 2.65 2.19 1.64
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t 3') Includes 0.5 % reduction for bypass leakage.
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TABLE 2 PARAMETERS USED IN FUEL HANDLING ACCIDENT ANALYSIS
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Accident in Accident in Fuel Storage Containment Pool Area
.(Auxiliary Buildina) L
, Core thermal power (MWt) 2831 2831 Time from' shutdown to accident (h) ' 100 100 Minimum waterdepth (ft) - 23 23 Damage to fuelassembly. All rods ruptured . All rods ruptured Activity release from assembly.
Kr-85 ' 30 % 30 %
' Other Noble Gases ,
10 % ' 10%
I-131- 12% - 12 %
Other lodines 10% , 10 %
Radial peaking factor ' l.7 1,7 Decontamination factor in water Iodine ~ - Elemental 133 133
- Organic '1 1 Noble gases 1 1 5 5
' Amount of mixing in building (ft')' 6.6 x 10 ' 1 x 10 s
- d Exhaust flow rate'(cfm) 5.35 x 10 12000 Isolation time (sec) 45 None Iodine filtration system None PRF Filter efficiency N/A 89.5 M*)
Atmospheric dilution factors (see FSAR (see FSAR y table 15B-2) table 15B-2) l i
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1 Accident in - Accident in Fuel Containment Storane Pool Area .
Site Bous&ry( w (REM)
Thyroid 12.1 21.6 Whole Body ' O.4 0.4 Low-Population Zone Dose (REM) '
Thyroid ' 4.5 7.9 Whole Body' O.1 0.1 4
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l' Joseph M. Farley Nuclear Plant I
Control Room, Penetration Room, and Containment Purge Filtration Systems and Radiation Monitoring Instrumentation r
Technical Specification Changes j Revised 10 CFR 50.92 Evaluntion
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I Enclosure 2 Joseph M. Farley Nuclear Plant Control Room, Penetration Room, and Containment Purge Filtration Systems and Radiation Monitoring Instrumentation Technical Specification Changes .
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.R. evised 10 CFR 50.92 Evaluation i
Pursuant to 10 CFR 50.92, SNC has evaluated the proposed amendments and has determined that I operation of the facility in accordance with the proposed amendments would not involve a significant hazards consideration. The basis for this determination is as follows:
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- l. The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. I The proposed changes to convert from ANSI N510-1980 to ASME N510-1989 for specific FNP filtration surveillance testing requirements and related changes do not affect the probability of any accident occurring. The consequencea of any accident will not be affected since the proposed changes will continue to ensure that appropriate and required surveillance testing for FNP filtration systems will be performed consistent with the revised accident analyses. The 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, remain within the guidelines of 10 CFR Part 100 and General Design Criterion 19 of Appendix A to 10 CFR Part 50. Relocating specific testing requirements to the FNP FS AR has no effect on the probability or consequences of any accident previously evaluated since required testing will continue to be performed.
Therefore, the proposed Technical Specification 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 difTerences 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 additional channels of radiation monitoring and verific.ttion of penetration room boundary integrity do not require the affected systems to be placed in configurations different from design. Thus, no new system design or testing configuration 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 testing requirements to the FSAR has no effect on the '. ossibility of creating a new or different kind of accident from any accident previously evaluated since it is an administrative change in nature.
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l Enclosure 2 Revised 10 CFR 50.92 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 I N510-1989 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 continue to meet the guidelines of 10 CFR Part 100 and GDC 19 of Appendix A to 10 CFR Part 50. The enhaaced testing of radiation monitoring instrumentation, the penetration room boundary negative pressure test, and the spent fuel room negative pressure test provide additional assurance that the acceptance criteria of the safety analyses and the resultant ,
margins of safety are not reduced. Relocating specific testing requirements to the FSAR has no effect I on the margin of plant safety since required testing will continue to be performed. Clarifying the 10 l hour mn with heaters on is consistent with the Improved Technical Specification 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 plant safety as defined in the Final Safety Analysis Report or the bases of the FNP technical specifications.
. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
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. Conclusion 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 previously 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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