ML20247F798
| ML20247F798 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 02/26/1998 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20247F730 | List:
|
| References | |
| NUDOCS 9805200050 | |
| Download: ML20247F798 (38) | |
Text
Attachment I to JPN-98-007 PROPOSED TECHNICAL SPECIFICATION CHANGES (JPTS-97-007)
New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 33 P PDR
Attachment V to JPN 98-007 SUPPORTING CALCULATIONS AND BASES FOR METEOROLOGY (JPTS-97-007) i New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 I DPR-59
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JAFNPP 6.19 POSTACCIDENT SAMPLING PROGRAM A program shall be established, implemented, and maintained which will ensure the capability to obtain and analy_e reactor coolant, radioactive iodines and particulate in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shallinclude the following: .
1 A) Training of personnel, B) Procedures for sampling and analysis, C) Provisions for maintenance of sampling and analysis 6.20 PRIMARY CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the Primary Containment ss required by 10 CFR 50.54 (o) and 10 CFR 50, Appendix J, Option B, as rnodified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment I Leak-Test Program", dated September 1995, as modified by the exception that Type C l testing of valves not isolable from the containment free air space may be accomplished j by pressurization in the reverse direction provided that testing in this manner provides i equivalent or more conservative results than testing in the accident direction. If l potential atmospheric leakage paths (e.g., valve stem packing) are not subjected to test pressure, the portions of the valve not exposed to test pressure shall be subjected to leakage rate measurement during regularly scheduled Type A testing. A list of these valves, the leakage rate measurement method, and the acceptance criteria, shall be contained in the Program.
A. The peak Primary Containment internal pressure for the design basis loss of coolant accident (P ), is 45 psig.
B. The maximum allowable Primary Containment leakage rate (L,), at P,, shall be 1.5% of primary containment air weight per day. l C. The leakage rate acceptance criteria are:
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- 1. Primary containment leakage rate acceptance criteria is 51.0 L,.
During unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 5 0.60 L, for the Type B and Type C tests and 5 0.75 L, for the Type A tests;
- 2. Airlock testing acceptance criteria are: i
- a. Overall airlock leakage rate is 5 0.05 L, when tested at 2 P,,
- b. For each door seal, leakage rate is 5120 scfd when tested at 2 !
P, .
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D. The provisions of Specification 4.0.B do not apply to the test frequencies specified l l in the Primary Containment Leakage Rate Testing Program. j l i E. The provisions of Specification 4.0.C are applicable to the Primary Containment i Leakage Rate Testing Program.
l Amendment No. 130,234, 258e N_-_____________-_ i
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l RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATIONS 1.0 DEFINITIONS A. Dose Eauivalent 1-131 The Dose Equivalent 1-131 is the concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131,1-132,1-133, i-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in International Commission on Radiological Protection Publication 30 (ICRP-30), " Limits for Intake by Workers" or in NRC Regulatory Guide 1.109, Revision 1, October 1977.
B. Instrument Channel Calibration See Appendix A Technical Specifications.
C. Instrument Channel Functional Test See Appendix A Technical Specifications.
D. Instrument Check See Appendix A Technical Specifications.
E. Loaic System Function Test See Appendix A Technical Specifications.
F. Member (s) of the Public Member (s) of the Public includes all persons who are not occupationally associated with the facilities on the NYPA/(NMPC) Niagara Mohawk Power Corporation site. This category does not include employees of the utilities, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plants.
G. Offaas Treatment System The Offgas Treatment System is the system designed and installed to: reduce radioactive gaseous effluents by collecting primary coolant system offgases from the main condenser; and, providing for delay of the offgas for the purpose of reducing the total radioactivity prior to release to the environment.
H. Offsite Dose Calculation Manual (ODCM)
The ODCM describes the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluents monitoring instrumentation alarm / trip set points and il the conduct of the environmental monitoring program.
- 1. Operable See Appendix A Technical Specifications.
Amendment No. 93, 1
Attichm::nt 11 t3 JPN-98-007 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES UPTS-97-007)
I i
New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 OPR-59 l
- - - - - - - - - - - - - - - - - - - - _ - - _ - - - _-__ a
1 Attachm:nt il to JPN-98-007 SAFETY EVALUATION FOR l PROPOSED TECHNICAL SPECIFICATION CHANGES l TO CLARIFY ALLOWED CONTAINMENT LEAKAGE RATE (L) (JPTS-97-007) '
l l
- l. DESCRIPTION OF THE PROPOSED CHANGES The proposed TS amendment changes the allowable containment leakage rate (L .) to 1.5 percent per day, and changes the assumed value of the filter efficiency for the Standby Gas Treatment System (SBGT) used in the calculations to determine dose l effects to onsite and offsite receptors during a Design Basis Loss of Coolant
{
l' Accident. Also, an ancillary change is proposed to correct conflicting information in l TS section 4.6.C, " Coolant Chemistry". The specific changes are as follows:
- 1. Page 139, section 4.6.C.1.d., change "These sampling requirements may be omitted whenever the equilibrium I-131 concentration in the reactor coolant is less than 0.007pCi/ml." to "These sampling requirements may be omitted whenever the equilibrium I-131 concentration in the reactor coolant is less j than .002 Ci/ml."
- 2. Page 140, section 4.6.C.1.e, change 0.007 pCi/ml to 0.002 pCi/ml.
l I
- 3. Page 191, section 3.7.C Bases. Replace the third full paragraph beginning with "The 99 percent efficiency ..." with "The analysis of the design basis loss-of-coolant accident assumed a charcoal filter efficiency of 90% for the SBGT system and a source term provided by GE based on NEDO-10871.
The assumed 90% is sufficient to prevent exceeding 10CFR100 guidelines for accidents analyzed. The charcoal and particulate filters are tested to an acceptance criteria of 99% efficiency with 1 % penetration. A heater maintains relative humidity below 70% in order to assure the efficient removal of methyliodine on the impregnated charcoal filters. Regulatory Guide 1.52 assigns a charcoal filter efficiency of 95% for 2 inch beds (as used in the SBGT system), thus assuming 90% efficiency in dose calculations and testing to 99% efficiency provide additional conservatism in analysis and operation."
- 4. Page 193, section 4.7 Bases for Primary Containment, paragraph 3. Change "The design basis accident leakage rate is 0.5 percent / day at 45 psig to "The
- design basis accident leakage rate is 1.5 percent / day at 45 psig."
- 5. Page 193, section 4.7 Bases for Primary Containment, paragraph 4. Change i
"... and the standby gas treatment system filter efficiency was 99% for halogens." to "... and the standby gas treatment system filter efficiency was 90% for halogens."
- 6. Page 258e, Administrative Controls, section 6.20 entitled " Primary Containment Leakage Rate Testing Program", change section 6.20.B, from "The maximum allowable Primary Containment leakage rate (L ), at P., shall be 0.5% of primary containment air weight per day" to "The maximum allowable Primary Containment leakage rate (L), at P., shall be 1.5% of primary containment air weight per day.
Page 1 of 11 1
Att: chm:nt ll to JPN-98-007 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES TO CLARIFY ALLOWED CONTAINMENT LEAKAGE RATE (L,) (JPTS-97-007)
- 7. Pages 285, References 20. Change the calculation revision number from Rev. O to Rev. 2.
- 8. Pages 285a, References 21. Change the calculation revision number from Rev. O to Rev.1.
- 9. Page 285a, add Reference 23, James A. FitzPatrick Calculation JAF-CALC-RAD-00007, Rev. 2, " Power Uprate Program - Onsite and Offsite Atmospheric Dispersion Factors". November 1997.
- 10. Page 1 of Appendix B, " Radiological Effluent Technical Specifications",
change wording in the definition of " Dose Equivalent 1-131" from "... Table ill of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites" o.- in Regulatory Guide 1.109, Reference 1, October 1977." to
" ..lCRP-30, " Limits for intake by Workers" or in Regulatory Guide 1.109, Reference 1, October 1977."
ll. PURPOSE OF THE PROtQSED CHANGES The proposed TS amendment changes the allowable containment leakage rate (L .) to 1.5 percent per day, and changes the assumed value of the filter efficiency for the Standby Gas Treatment System (SBGT) used in the calculations to determine dose effects to onsite and offsite receptors during a Design BaW, Loss of Coolant Accident. Also, an ancillary change is proposed to correct conflicting information in TS section 4.6.C, " Coolant Chemistry".
The change in the allowed containment leakage rata is necessary to correct an NRC identified discrepancy between the James A. FitzPatrick (JAF) Final Safety Analysis Report (FSAR) and the original NRC Safety Evaluation Report for JAF (NRC inspection report 82-04.) Attachment IV contains history and background information pertaining to the identified discrepancy. In addition to this, the 4 Authority requests this amendment because the misapplication of the ILRT and LLRT I acceptance criterion, design leakage rate of (L ) at 0.5% wt/ day versus the value allowed by Appendix J, L , has resulted in unnecessary exposure to plant personnel and expenditure of resources by the Authority.
The assumed standby gas treatment charcoal filter efficiency used in the calculations that predict the radiologicalimpact to onsite and offsite personnel following a design basis accident was changed from 99% to 90% efficiency for halogens. The proposed change on page 191, section 3.7.C Bases, replaces the thkd full paragraph beginning with "The 99 percent efficiency ..." with "The analysis of the design basis loss-of-coolant accident assumed a charcoal filter efficiency of 90% for the SBGT system and a source term provided by GE based on NEDO-10871. The assumed 90% is sufficient to prevent exceeding 10CFR100 guidelines for accidents analyzed. The charcoal and particulate filters are tested to an acceptance criteria of 99% efficiency with 1 % penetration. A heater maintains l
Page 2 of 11 j
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l Att: chm:nt ll t3 JPN-98-007 1
SAFETY EVALUATION FOR j PROPOSED TECHNICAL SPECIFICATION CHANGES j TO CLARIFY ALLOWED CONTAINMENT LEAKAGE RATE (L,) (JPTS-97-007) relative humidity below 70% in order to assure the efficient removal of methyl ,
iodine on the impregnated charcoal filters." This change is necessary to be l consistent with NRC Regulatory Guide 1.52, Rev 2, March 1978. JAFNPP I previously took exception to the Regulatory Guide in Technical Specification Amendment 239 (Power Uprate). To support Amendment 239, the Authority committed to the acceptance criteria that is required for 4 inch filters, per the Regulatory Guide, of .175% percent penetration in order to assume 99% efficiency I of the charcoal filters. JAFNPP has 2 inch charcoal beds in the Standby Gas Treatment System. NRC Regulatory Guide 1.52 table 2 assigns an activated charcoal efficiency of 95% for a 2 inch charcoal bed, with a test for methyliodide penetration of less than 1 %. The revised calculations assume a more conservative charcoal efficiency than the Regulatory Guide 1.52 assigned value for 2 inch beds.
The fission product source term methodology, TID-14844, referenced on page 191, paragraph 3 was changed to a GE provided source term that was primarily based on NEDO-10871. This change to the TS should have been made during the Power Uprate submittal, TS amendment 239. The calculations used to support the Uprate submittal used the GE source term versus the previous one based on TID-14844 methodology. The calculations that reference the new source term were incorporated into TS as part of the power uprate submittal and are currently listed as reference 20 and 21.
Technical Specification section 4.6.C. " Coolant Chemistry" subsection 1.d. states that sampling requirements may be ornitted whenever the equilibrium I-i31 concentration in the reactor is less than .007 Ci/mi, and subsection 1.e states
"...with a, c, and d above indicate a total iodine concentration in excess of 0.007 Ci/mi, a quantative determination shall be made for 1-131 and l-133." A change in the value of .007 vCi/ml to .002 Ci/mlis necessary to agree with a statement contained in the bases section of 4.6.C. " Coolant Chemistry". The bases section states " Analysis is required whenever the 1-131 concentration is within a factor of 100 of its allowable equilibrium value." The allowable equilibrium value per TS section 3.6.C. is 0.2 Ci/gm of dose equivalent 1-131. The change from .007 Ci/mi to .002 Ci/miis minor, more conservative, and will correct the inconsistency between the technical specification and its basis.
TS references 20 and 21 have been revised to conservatively assume a Standby Gas Treatment system (SBGT) charcoal efficiency of 90% and to incorporate recently revised atmospheric dispersion factors (added reference 23 to TS).
The proposed change on page 1 of Appendix B, " Radiological Effluent Technical Specifications", to change wording in the definition of " Dose Equivalent I-131" from
"... Table 111 of TID-14844, "CaLulation of Distance Factors for Power and Test Reactor Sites" or in Regulatory Guide 1.109, Reference 1, October 1977." to
"...lCRP-30, " Limits for Intake by Workers" or in Regulatory Guide 1.109, Reference 1, October 1977." is necessary to reflect the thyroid dose conversion factors used in the revised calculations (reference 20 and 21 of TS).
Page 3 of 11
Attachmsnt il to JPN-98-007 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES TO CLARIFY ALLOWED CONTAINMENT LEAKAGE RATE (L,) (JPTS-97-007) l I The Authority has evaluated the proposed TS Amendment and determined that it does not represent a significant hazards consideration as defined in 10CFR50.92.
and meets the criterion set forth in 10CFR51.22 for " categorical exclusion."
lli. SAFETY IMPLICATIONS OF THE PROPOSED CHANGEG
- 1. Changing the Value of L; Proposed Specification 6.20 specifies the allowable containment leakage rate (L.) as 1.5 percent of primary containment air weight per day. This is a correction to the value of L, for the FitzPatrick plant. This correction was previousiv submitted as part of a proposed TS amendment (Reference 2), and subsequently withdrawn (Reference 3) because additional evaluation was required to quantify the effects of a 1.5 percent per day leakage rate on safety-related equipment located in the reactor building. The Authority has reviewed the environmental qualification of safety-related equipment in the react'ar building and identified two component types that required further evaluation as a result of this correction. Appropriate action has been taken to qualify these components to support an allowable leakage rate of 1.5 %wt/ day.
The original TS defined the Type A acceptance criteria as less than 0.75 L, and not greater than the design leakage rate, L, (0.5%/ day). This acceptance criteria has be en retained and is currently the basis for the acceptance criteria of 0.5% ut/ day contained in the Primary Containment Leak Rate Testing Program (Reference 1). This acceptance criteria was contained in the original TS and was written to support the pre-operational test. The original TS SR is consistent with current pre-operational leakage rate test requirements of Option A, paragraph Ill.A.4(b)(2), and conservative with respect to the retest leakage rate requirement of Option A, paragraph lli.A.5(b)(2), which defines the acceptance criteria as less than 0.75L . For purpotes of establishing Type A, B and C leakage test acceptance criteria, the allowable containment leakage rate has been limited to 0.5 weight percent of the contained air volume per day so as not to conflict with the original SR. This interpretation is overly conservative with respect to the TS Bases, and the current analytical basis.
The FitzPatrick accident analyses assume an allowable leakage rate (L.) of 1.5 percent weight per day. This limitation on containment leakage rate ensures that totalleakage will not exceed the value assumed in the accident analyses at the peak accident pressure (P,) of 45 psig. The margin of safety for the off-site dose consequences of postulated accidents directly related to the containment leakage rate is maintained by meeting the 1.0L , (1.5%
wt/ day) acceptance criteria stated in Specification 6.20.
Page 4 of 11
Att: chm:nt il ta JPN-98-007 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES TO CLARIFY ALLOWED CONTAINMENT LEAKAGE RATE (L,) (JPTS-97 007)
The effects of this change ore: 1) The value of the As-Left Type A test leakage criteria of 0.75 L,is 1.125 percent per day; 2) The value of the combined Type B and C test leakage acceptance criteria of 0.6 L,is 0.9 percent per day; and 3) The value of the "As-found" Type A test acceptance criteria is 1.5 percent per day (L.). The value of 1.5 percent per day is consistent with the accident analyses and Option B, and does not constitute an increase in the allowable leakage rates as analyzed in the UFSAR.
Therefore, this change does not adversely impact plant safety.
From a risk perspective evaluation, past studies show that overall reactor I accident risks are not sensitive to variations in containment leakage rate.
This is because reactor accident risks are dominated by accident scenarios in I which the containment fails or is bypassed. Such scenarios, even though f they are of very low probability, dominate the predicted accident risks due to their high consequences. FitzPatrick Individual Plant Examination (IPE) results are consistent with these past technical studies.
Certain NRC sponsored studies (References 4 and 5) indicate that overall plant risk is not sensitive to changes in containment leak rates. As can be seen on Table 2 the incremental risk from leakage in the range of 1 % to 10%
per day is small. FitzPatrick can be compared with Peach Bottom, both are BWR 4 plants with MARK I containments.
- 2. Chance in the Assumed Value for SBGT Svstem Filter Efficiencv The assumed standby gas treatment charcoal filter efficiency used in the calculations that predict the radiological impact to onsite and offsite receptors following a design basis accident was changed from 99% to 90%
efficiency for halogens. This change is necessary to be consistent with NRC Regulatory Guide 1.52, Rev 2, March 1978. NRC Regulatory Guide 1.52 table 2 assigns an activated charcoal efficiency of 95% for a 2 inch charcoal bed, with a test for methyl iodide penetration of less than 1 %. JAFNPP previously took exception to this in Technical Specification Amendment 239 (Power Uprate) and committed to testing the charcoal filters to an acceptance criteria of .175% penetration in order to assurne 99% efficiency of the charcoal filters. The JAFNPP has 2 inch charcoal beds in the staridby gas treatment system. The revised calculations assume a value for charcoal efficiency that is conservative compared to the assigned values of Regulatory Guide 1.52 for charcoal efficiency for 2 inch beds and are conservative for the purpose of calculating dose to onsite and offsite receptors.
l Page 5 of 11
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l Att chm:nt ll to JPN-98-007 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES TO CLARIFY ALLOWED CONTAINMENT LEAKAGE RATE (L,) (JPTS-97-007)
- 3. Changing Limit in Bases section 4.GS " Coolant Chemisity" Technical Specification section 4.6.C. " Coolant Chemistry" subsection 1.d.
states that sampling requirements may be omitted whenever the equilibrium I-131 concentration in the reactor is less than .007 Ci/ml. A change in the value of .007 Ci/ml to .002 Ci/mlis necessary to agree with a statement contained in the bases section of 4.6.C. " Coolant Chemistry". The bases section states " Analysis is required whenever the I-131 concentration is within a factor of 100 of its allowable equilibrium value." The allowable equilibrium value per TS section 3.6.C. is 0.2 pCi/gm of dose equivalent I-131. This change is minor and conservative with respect to the current l
value and is consistent with TS bases.
IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION The Authority has evaluated the proposed TS Amendment and determined that it does not represent a significant hazards consideration. Based on the criteria for defining a significant hazards consideration established in 10CFR50.92, operation of the James A. Fit? Patrick Nuclear Power Plant in accordance with the proposed amendment will not:
- 1) Involve a significant increase in the probability or consequences of an accident previously evaluated because:
The proposed changes do not involve a change to the design or operation of the plant. The systems affected by this proposed TS change are not assumed in any safety analyses to initiate any accident sequence. Therefore, the probability of any accident previously evaluated is not increased by this -
proposed TS change. The change in the allowable containment leakage rate (L.) is consistent with the accident analyses. The assumption of only 90%
SBGT filter efficiency is conservative with actual system performance and is consistent with Regulatory Guide 1.52. There is no significant change to the consequences of an accident previously evaluated because maintaining containment leakage within limits assumed in the accident analyses ensures that the dose consequences resulting from an accident are not increased. ,
The calculated doses with the decreased SBGT system charcoal efficiency '
for design basis accidents are marginally increased but still meet, and are well below, the dose acceptance criteria of 10CFR100, the SRP, and GDC 19 of Appendix A to 10CFR50. The proposed TS changes maintain an
]
1 equivalent level of reliability and availability for all affected systems. The .l ability of the affected systems associated with maintaining leak rate integrity I to perform their intended function is unaffected by the proposed TS changes. Implementation of these changes will provide continued assurance that specified parameters associated with containment integrity will remain 1 within acceptance limits, and as such, will not significantly increase the I consequences of a previously evaluated accident. The change in the value of
]
.007 Ci/ml to .002 Ci/ml in section of 4.6.C. " Coolant Chemistry" is a Page 6 of 11
l Att: chm nt il t2 JPN-98-007 l SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES TO CLARIFY ALLOWED CONTAINMENT LEAKAGE RATE (L ) (JPTS-97-007) minor editorial change, is more conservative, and will correct the inconsistency between the technical specification and its basis and as such, will not significantly increase the consequences of a previously evaluated accident .
- 2) Create the possibility of a new or different kind of accident from any accident previously evaluated because:
The proposed amendment changes the allowed containment leakage rate to 1.5%, changes the assumed value for SBGT system charcoal filter efficiency, and changes a specification in section of 4.6.C. " Coolant Chemistry" from the value of .007 Ci/ml to .002 Ci/mt. No new accident modes are created by clarifying the numerical value of the allowable containment leakage rate (L.) or changing the assumed value for the SBGT system charcoal filter efficiency. No safety-related equipment or safety functions are altered, or adversely affected, as a result of these changes. The proposed changes will not introduce failure mechanisms beyond those already considered in the current plant saf aty analyses. Changing the allowable leakage rate, the assumed value for the efficiency of the SBGT system charcoal filter, and the specification in the bases section of 4.6.C. " Coolant Chemistry" does not contribute to the possibility of a new or different kind of accident or malfunction from those previously analyzed.
- 3) Involve a significant reduction in the margin of safety because:
The proposed amendment changes the allowed containment leakage rate to 1.5%, changes,the assumed value for SBGT system charcoal filter efficiency, and changes a specification in section of 4.6.C. " Coolant Chemistry" from the value of .007 Ci/ml to .002 Ci/ml. The design of the FitzPatrick plant is not changed. The methodology for test performance is unchanged and Type A, B and C tests will continue to be performed at 1.P., The value of L, specified in proposed specification 6.20 is consistent with the accident analyses, therefore, the dose consequences of any esalyzed accidents are not increased as a result of this change. The calculated doses as a result of the decrease in the assumed efficiency of the SBGT system charcoal filters for design basis accidents are marginally increased but still meet, and are well below, the dose acceptance criteria of 10CFR100, the SRP, and GDC 19 of Appendix A to 10CFR50. The change in the specification in section 4.6.C. " Coolant Chemistry" from .007 pCi/ml to .002 Ci/miis a minor editorial change, is more conservative, and will correct the inconsistency between the technical specification and its basis. Therefore, the proposed changes provide continued assurance of the leak tightness of the containment and conservatively assume SBGT system charcoal filter efficiency for the purpose of dose calculations for design basis accidents without adversely affecting the public health and safety and, as such, will not involve a significant reduction in the margin of safety.
Page 7 of 11
Attichmint 11 ts JPN-98-007 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES TO CLARIFY ALLOWED CONTAINMENT LEAKAGE RATE (L,) (JPTS-97-007)
This proposed amendment does not involve a significant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings or a significant relaxation of the bases for the limiting conditions for operations. Therefore, based on the criteria established in 10CFR50.92(c), the proposed change does not constitute a significant hazards consideration.
V. ENALUATION OF CATEGORICAL EXCLUSION PURSUANT TO 10CFR 51.22fcil9) 10CFR51.21 requires performance of an Environmental Assessment for alllicensing and regulatory actions, except for those identified in 10CFR51.20 requiring an Environmental Impact Statement, and those identified in 10CFR51.22 as " Categorical Exclusions." The section also contains the caveat that the Commission may, in special circumstances, prepare an environmental assessment on an action covered by a categorical exclusion.
A Categorical Exclusion is defined as a category of actions which do not individually or cumulatively have a significant effect on the human environment and which the Commission has found to have no effect in accordance with the procedures set out in Part 51,22; and for which, therefore neither an environmental assessment nor an environmental impact statement are required.
Stated in part in 10CFR51.22(C)(9), License Amendments which change a requirement with respect to installation or use of a facility component located within the Restricted Area as defined in Part 20, or which change an inspection or surveillance requirement, are considered as categorical exclusions provided that:
- 1) The Amendment involves no significant hazards consideration.
As concluded above, the clarification of the value of L , and changing a assumed value of the SBGT system charcoal filter efficiency does not involve a significant hazards consideration.
- 2) No significant change in the types, or significant increase in the amount of effluent that may be released off-site.
Changing the value of L, does not change the types of materials released during and following an accident. The design containment leakage is not changed by this clarification, therefore this change does not affect the realistic assumptions used in the FitzPatrick Final Environmental Statement. The value of 1.5 percent per day for L,is consistent with the TS bases and the FSAR, and is the value used in the accident analyses, therefore the 1.5% change has no impact on the calculated post-accident doses. The calculated doses do increase as a result of changing the assumed charcoal filter efficiency from 99% to 90% (reference 6) >
to offsite receptors during a design basis loss of cooling accident, however are well below those required by 10CFR100. The new calculated doses at the site boundary and LPZ are < 25% of the 10CFR100 criteria.
Page 8 of 11
i l
Atttchment 11 to JPN-98-007 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES TO CLARIFY ALLOWED CONTAINMENT LEAKAGE RATE (L.) (JPTS-97-007) l l 3) No significant increase in individual or cumulative occupational radiation exposure.
The proposal makes no design changes to the primary containment, only changes the Appendix J leakage rate testing acceptance criteria. The design leakage rate of the containment is not affected. The value of L. is changed to make it consistent with applicable regulations and the Plant Safety Analysis.
The calculated doses (reference 6 and 7) to personnel within the protected area during a design basis accident, with the change in the assumed value for the efficiency of the SBGT system charcoal filters, are below those required by 10CFR20 and 10CFR100.
VI. CONCLUSION
- 1) The design leakage of the FitzPatrick primary containment is 0.5 percent per day. This value is consistent with the General Electric specified design leakage for Mark i BWR containments.
- 2) The current acceptance criteria in the Containment Leakage Rate Program is consistent with the Appendix J Paragraph Ill.A.4(b)(2) acceptance criteria for the initial preoperational test that required that primary containment leakage be verified less than design.
- 3) The FitzPatrick accident analyses used a value of 1.5 percent per day as the maximum allowable primary containment leakage rate for calculation of accident doses for 10CFR100 purposes.
- 4) Based on the discussions above, the change in the containment allowed leakage rate in the TS will not decrease the effectiveness of the containment or containment leakage rate testing. Operating limitations will continue to be imposed, and required surveillance will continue to be performed in accordance with Technical Specifications, written procedures and instructions auditable by the NRC. The assumptions in the FitzPatrick licensing bases are not invalidated by the proposed Technical Specification change.
- 5) Based on the discussions above the calculated doses as a result of the decrease in the assumed efficiency of the SBGT system charcoal filters for design basis accidents are marginally increased but st:ll meet and are well below the dose acceptance criteria of 10CFR20 and 100, the SRP, and GDC 19 of Appendix A to 10CFR50.
- 5) The Flant Operating Review Committee (PORC) and the Safety Review Committee (SRC) have reviewed these proposed changes to the Technical Specifications and have concluded that they do not involve an unreviewed safety question, or a significant hazards consideration, and will not endanger the health and safety of the public.
Page 9 of 11
AttachmInt il to JPN-98-007 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES TO CLARIFY ALLOWED CONTAINMENT LEAKAGE RATE (L ) (JPTS-97-007)
- 6) By submitting a TS amendment application, the Authority has taken the proper actions to resolve the L. discrepancy that exists between the FitzPatrick FSAR and the NRC SER .
- 7) Implementation of the proposed changes will not adversely affect the ALARA or Fire Protection Programs at the FitzPatrick plant, nor will the changes affect the, environment.
Vll. BEFERENCES
- 1. James A. FitzPatrick NPP " Primary Containment Leakage Rate Testing Program Plan", JAF-RPT-PC-02342,
- 2. NYPA letter, J.C. Brons to the NRC (JPN-90-008), " Proposed Change to the Technical Specifications Regarding Containment Leak Rate Testing Requirements (JPTS-84-012)," dated January 16,1990,
- 3. NYPA letter, R.E. Beedle to the NRC (JPN-92-016), " Withdrawal of Amendment Application (JPTS-84-012)," dated March 31,1992.
- 4. NUREG/CR-4330, " Review of Light Water Reactor Regulatory Requirements, Assessment of Selected Regulatory Requirements that may have Marginal Importance to Risk - Reactor Containment Leakage Rates - Main Steam isolation Valve Leakage Control Systems - Fuel Design Safety Reviews," Volume 2, dated June 1,1986.
- 5. NUREG-1150, " Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, Final Summary Report," dated December 1,1990.
- 6. JAF-CALC-RAD-0048, Revision 1, " Power Uprate Project - Radiological Irrpact at Onsite and Offsite Outdoor Receptors Following Design-Basis Accidents."
- 7. JAF-CALC-RAD-0042, Revision 2, " Control Room Radiological Habitability Under Power Uprate Conditions and CREVASS Reconfiguration."
Page 10 of 11
Att chm:nt ll to JPN-98-007 SAFETY EVALUATION FOR PROPOSED TECHNICAL SPECIFICATION CHANGES TO CLARIFY ALLOWED CONTAINMENT LEAKAGE RATE (L.) (JPTS-97-007)
Table 1 Post Core Damage (Level 3) Comparison of Results Population Dose, person-rem / reactor year Peach Bottom Grand Gulf Leak Rate % / day NUREG/ NUREG- NUREGl NUREG CR- 1150t2' CR- 1150 4330"I 4330 0.5 151 28.3 250 5.66 1 151 250 5 153 28.3 254 5.67 _
10 153 254 50 174 28.4 288 5.81 100 174 288 I
._ 1 1See Reference 4 2See Reference 5 Page 11 of 11
Att:chmint til to JPN-98-007 l
.. MARKUP OF TECHNICAL SPECIFICATION PAGES l (JPTS 97-007) l l
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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59
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6.19 POSTACCIDENT SA"MPLING PROGRAM A program shall be established, implemented, and maintained which will ensure the capability to obtain and analyze reactor ecolant, radioactive iodines and particulate in plant gaseous affluents, and containment atmosphere samples under accident conditions. The program shallinclude the following: A) Training of personnel, , B) Procedures for sampling and analysis, C) Provisions for maintenance of sampling and analysis 6.20 PRIMARY CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakaos rate testing of the Primary Contamment as required by 10 CFR 50.54 (c) and 10 CFR 50, Appendix J. Option B, /)I ' as modified by approved exemptions. This program shall be in accordance with the , guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program *, dated September 1995, as modified by the exception that Type C testing of valves not isolable from the containment tree air space may be accomplished by pressurization in the reverse direction provided that testing in this manner provides equivalent or more conservative results than testing in the accident direction, if potential atmospheric leakage paths (e.g., valve stem paciong) are not subjected to test . pressure, the portions of the valve not exposed to test pressure shall be subjected to leakage rate measurement during regularly scheduled Type A testing. A list of these valves, the leakage rate measurement method, and the acceptance criteria, shall be contained in the Program. ; f A. The peak Primary Containment intemal pressure for the design basis loss of coolant accident (P ), is 45 psig.
' um allowable Primary Containment leakage rate (L.), at P., shall be l.(% . B. I b of primary containment air weight per day.
C. The leakage rate acceptance criteria are: I
- 1. Primary containment leakage rate acceptance criteria is s 1.0 L., l i
During unit startup following testing in accordance with this program, the leakage rate a tance enteria are s 0.60 L, for the Type B and Type l C tests and s 0.7 L, for the Type A tests: 4
- 2. Airlock testing acceptance criteria are: l
- a. Overall airlock leakage rate is s 0.05 L, when tested at a P.,
- b. For each door seal, leakage rate is s 120 scfd when tested at 2 .
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.^ ' when tested at a 25 psig. D. The provisions of Specification 4.0.B do not apply to the test frequencies specified / in the Pnmary Containment Leakage Rate Testing Program. E. The provisions of Specification 4.0.C are applicable to the Primary Containment i Leakage Rate Testing Program. Amendment No. O sn . 258e
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Q\ 6 RADI0 LOGICAL EFFLUENT TECHNICAL SPECIFICATIONS 1.0 DETINIT!0NS A. Dose Equivalent I-131 The Dose Equivalent 1-131- is the concentration of I-131 (nicrocu-rie/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131.1-132. I-133. I-134 and I-135 actually present. The thyroid dose c3 aversion factors used for this p calculation shall be those listed inf Table u1 at m -iiii1 @ u-j
-(Tat taa af nf atance Factora__ for Power and Test *==eter Sicaa">or in NRC Regulatory Guide 1.109. Rev'ision 1. October 1977.
- 3. Instrument channel Calibration See Appendix A Technical Specifications. ,
C. Instrument Channel Funettonal Test , See Appendix A Technical Specifications. g4 ggg. D. Instrument check /f y .,, g g gf. See Appendix A Technical Specifications. ,
,g g g g 'J E. Lonic System Function Test 0 je, d f,-
see Appendix A Technical Specifications. T. Member (s) of the Public Member (s) of the Public includes all persons who are not occupation-ally associated with the facilities on the NTPA/(NMPC) Niagara Mohawk Power Corporation site. This category does not include employees of. the utilities, its contractors or vendors. Also excluded from this * ( category are persons who enter the site to service equipment or to j aske deliveries. This category does include persons who use portions of the site for recreational, occupational, u other purposes not associated with the plants. G. _offsas Treataset Systen The Offgas Treatammt System is the system designed and installed to: reduce radioactive gaseous affluents by collecting primary coolant system offgases from the main condenser and, providing for delay of the effgas for the purpose of reducing the total radioactivity prior to release to the environment. E. Offsite Dose Calculation Manual (CDCM) The CDCN describes the methodology and parameters to be used in the calculation of offsite doses due to radioactive gaseous and liquid l effluents and in the cr iculation of gaseous and liquid .e'ffluents l monitoring instrumentation alara/ trip set points and in the conduct of the environmental sonitoring program. I. Operable See Appendix A Technical Specifications. Amendment No. 93 1 i l
Attachment til t2 JPN-98-007 MARKUP OF TECHNICAL SPECIFICATION PAGES Inserts
- 10. sert A The analysis of the design basis loss-of-coolant accident assumed a charcoal filter efficiency of 90% for the SBGT system and a source term provided by GE based on NEDO-10871. The assumed 90% is sufficient to prevent exceeding 10CFR100 guidelines for accidents analyzed.
The charcoal and particulate filters are tested to an acceptance criteria of 99% efficiency with 1% penetration. A heater maintains relat.ve humidity below 70% in order to assure the efficient removal of methyliodine on the impregnated charcoal filters. Regulatory Guide.1.52 assigns a charcoal filter efficiency of 95% for 2 inch beds (as used in the SBGT system), thus assuming 90% efficiency in dose calculations and testing to 99% efficiency provide additional conservatism in analysis and operation. l l
f 1 Att:chmsnt IV to JPN-98-007 HISTORY AND BACKGROUND INFORMATION PERTAINING TO THE DISCREPANCY BETWEEN THE FSAR AND NRC SAFETY EVALUATION REPORT All OWFD LEAKAGE RATES l (JPTS-97-007) i l l New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59
Attachmsnt IV to JPN-98-007 HISTORY AND BACKGROUND INFORMATION PERTAINING TO THE DISCREPANCY BETWEEN THE FSAR AND NRC SAFETY EVALUATION REPORT ALLOWED LEAKAGE RATES
Background
10 CFR 50 Appendix J Option A defines allowable and design primary containment leakage rates as follows: L. (% per 24 hours) means the maximum allowable leakage rate at pressure P, as specified for preoperational tests in the technical spe.c4ications or associated bases, and as specified for periodic tests in the operating license. Le (% per 24 hours) means the design leakage rate at pressure, P., as specified in the TS or associated bases. The JAF Technical Specifications currently do not define the values of L, or La in the TS acceptance criteria, nor are they contained in Section 1.0, Definitions. However, the associated Bases (Section 4.7) states that " Design basis accidents were evaluated as discussed in Section 14.6 of the FSAR and the power uprate safety evaluation, Reference 18. The whole body and thyroid doses in the control room, low population zone (LPZ) and site boundary meet the requirements of 10 CFR Parts 50 and 100. The technical support center (TSC), not designed to these licensing bases, was also analyzed. The whole body and thyroid dose acceptance criteria used for the main control room are met for the TSC when initial access to the TSC and occupancy of certain areas in the TSC is restricted by administrative control. The LOCA dose evaluations, References 19,20, and 21, assumeu: the primary containment leak rate was 1.5 volume percent per day; source term releases were in accordance with TID-14844 and Regulatory Guide 1.3, and were consistent with the Standard Review Plan; and the standby gas treatment system filter efficiency was 99% for halogens. These doses are also based on the Amendment No. 239." It can be concluded, based on this statement, that L, for the FitzPatrick plant is 1.5 percent / day. The original TS defined the Type A acceptance criteria as less than 0.75 L,and not greater than the design leakage rate, L,(0.5%/ day). This acceptance criteria was written to support the pre-operational test to verify the design leakage rate of less than 0.5%/ day. This acceptance criteria has been retained and is currently the bases for the acceptance criteria of 0.5% wt/ day contained in the Primary Containment Leak Rate Testing Program. The original TS SR is consistent with the current pre-operationalleakage rate test requirements of Option A, paragraph Ill.A.4(b)(2), and conservative with respect to the retest leakage rate requirement of Option A, paragraph Ill.A.5(b)(2), which defines the acceptance criteria as less than 0.75L,. For purposes of i establishing Type A, B and C leakage test acceptance criteria, the allowable containment leakage ! rate for JAFNPP, has been limited to 0.5 weight percent of the contained air volume per day so as not to conflict with the original SR. ) l l i l Page 1 of 6 l i
Attachment IV to JPN-98-007 l l HISTORY AND BACKGROUND INFORMATION PERTAINING TO THE DISCREPANCY BETWEEN THE FSAR AND NRC SAFETY EVALUATION REPORT ALLOWED LEAKAGE RATES Allowable Primary Containment Leakage Rate The following is a history pertaining to the allowable primary containment leakage rate for the James A. FitzPatrick Nuclear Power Plant: 1969 The PSAR defined the design primary containment leakage as 0.5 percent per day, in the PSAR LOCA analysis, leakage rates were assumed to be 0.5% of the contained free volume per day at 25 psig using the turbulence (rough passage) equation for interpolating higher and lower pressures. 3/4/70 NRC Construction Permit SER Section 6.1 stated the design leakage rate of the containment as 0.5 percent per day. For LOCA analysis purposes the containment was assumed to leak at a constant rate of 0.6 percent per day for the duration of the accident without consideration of the effects of decreasing pressure during the post-accident interval. 6/71 The Original FSAR submitted to the NRC defined the value of the design containment leakage as 0.5 percent per day, and the value of allowable leakage rate as 1.5 percent per day. These numbers appear in various parts of the original FSAR. Section 5.2.4.4 states:
"For purposes of evaluating containment leakage in accident analyses, it was assumed that the primary containment has a leakage rate of 1.5 percent per day. For accident analysis purposes, this procedure is conservative since containment is designed to preclude leakage greater than 0.5 percent per day at the containment design pressure of 56 psi gage."
The Section 14.6.3.5 discussion regarding fission product release to the secondary containment following a LOCA states:
"The primary containment fission product release to the secondary containment was calculated assuming that the primary containment leak is i 1.5 percent of the contained free volume per 24 hrs at 25 psi gage, and using the turbulence (rough passage) equation for interpolation to higher and lower pressures. The long-term primary containment pressure is shown on figure 14.6-8 and was calculated based on the design leak rate of 0.5 percent per day of containment volume."
These statements show that the design containment leakage rate (Lo ) was 0.5% volume per day, and the maximum allowable leakage rate (L.) was 1.5% volume per day. The most recent updated FSAR is consistent with these statements. Page 2 of 6
l Attachment IV to JPM-98-007 HISTORY AND BACKGROUND INFORMATION PERTAINING TO THE DISCREPANCY BETWEEN THE FSAR AND NRC SAFETY EVALUATION REPORT ALLOWED LEAKAGE RATES i 11/29/71 In Question 5.10 (FSAR questions and answers) the NRC requested information regarding the primary containment leakage rate testing program be provided to ! ensure compliance with the proposed Appendix J. The PASNY response defined "L, = 1.5% per 24 hours," and stated that this information supplemented the j information in Appendix B (later the TS) Sections 3.7 and 4.7. 11/2/72 The NRC SER for the FitzPatrick plant summarized the results of the NRC staff review of the information and data submitted by PASNY. With regard to design containment leakage, the SER in Section 5.2 states that design containment leakage rate is 0.5 percent per day at the design pressure. This is consistent with the FitzPatrick FSAR. With regard to the Containment Atmosphere Dilution (CAD) system, Section 6.4 of the SER evaluates the rise in containment pressure post-LOCA, due to use of the CAD system, at both a zero containment leakage rate and an assumed leakage rate of 1.5 volume percent per day. Section 10.1 Accident Analysis states that "we (the NRC) performed conservative analysis of these design basis accidents to assess the adequacy of the engineering safety features to control and minimize the possible escape of fission products from the facility." In Section 10.2, Loss of Coolant Accident, the NRC states that "the primary containment was assumed to leak at a constant rate of 0.5 percent of its volume per day at accident conditions for the 30 day accident duration without consideration of the mitigating effect of decreasing pressure during the accident interval." The inconsistencies within the SER, and the contradiction between the FSAR and the NRC evaluation of accident consequences, were not identified as unresolved matters in the conclusion (Section 18.0) of the SER. Documentation between PASNY and the NRC to clarify these inconsistencies and contradictions could not be found. 3/4/82 NRC Inspection Report 82-04 identified the discrepancy between the FSAR and the NRC SER, and stated the following: J l j "Further examination of the FSAR and NRC independent accident analysis performed in 1972 revealed discrepancies in assumed values for L, and results of the LOCA analysis dose calculations. The inspector verified that , both calculation results satisfied the limits specified in 10 CFR 100. However the inspector informed the licensee that this matter would receive further NRC review." This issue was considered unresolved (333/82-04-02). Page 3 of 6 i
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Attachment IV to JPN-98-007 HISTORY AND BACKGROUND INFORMATION PERTAINING TO THE DISCREPANCY BEIWEEN THE FSAR AND NRC SAFETY EVALUATION REPORT ALLOWED LEAKAGE RATES 1 l l l 5/17/88 Inspection Report 88-05 closed Unresolved item 82-04-02. The Report acknowledged that JAF was procedurally meeting 10 CFR 50 Appendix J requirements in this area, however, the TS and Bases were unclear. It also acknowledged that the Authority had begun a proposed TS amendment (JPTS l 12) to resolve this issue. 1/16/90 Proposed TS amendment JPTS-84-12 was submitted (JPN-90-008) to resolve various Appendix J testing issues, including the clarification of L,. 3/31/92 The proposed TS amendment submitted on 1/16/90 was withdrawn (JPN-92-016) because of questions regarding qualification of certain safety-related components in the reactor building with an L, of 1.5% per day. Standbv Gas Treatment System Filter Efficiencv Original FSAR Review Section 5.3.3.4, " Standby Gas Treatment System," states that each HEPA filter is designed to be capable of removing at least 99.97 percent of the 0.30 micron particles which impinge on the filter, however, credit is taken only for a 90 percent removal capability. Section 14.6.3.6, LOCA Analysis Fission Product Release to the Environment, contains a discussion regarding charcoal filter efficiencies which presents experimental data that these filters have demonstrated removal capabilities in the range of greater than 99.9%. The section concludes by stating that "Thus, with this design the assumption of only 90 percent filter efficiency for the removal of inorganic and organic halogens by the standby gas treatment system is conservative by orders of magnitude 10t" Section 14.6.4.5, Refueling Accident Fission Product Release to the Environment, states that the amount of fission products released from the main stack is calculated assuming that the efficiency of the filters is only 90 percent. AEC Question 14.6 of 11/29/71 stated that the AEC calculations assume a 90% efficiency for removal of halogens by single filters, and 95% efficiency applied only where there are two such filters in series. PASNY was requested to provide data to demonstrate the equivalence of the JAF filter system to one having two filters in series. The response to this question stated that "We (PASNY) have revised the accident dose rate calculations using the AEC assumption of 90 percent efficiency for removal of halogens by a single filter." l 1 AEC Question 11.6 of 1/12/72 requested additional information regarding radiation doses in the control room as a result of the design basis accidents. Regarding Standby Gas Treatment System filter efficiency the response states that the efficiency for removal of airborne halogen activity is taken to be 90%. I 1 Page 4 of 6
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Attschmsnt IV to JPN-98-007 HISTORY AND BACKGROUND INFORMATION PERTAINING TO THE DISCREPANCY BETWEEN THE FSAR AND NRC SAFETY EVALUATION REPORT ALLOWED LEAKAGE RATES The NRC SER dated 11/2/72 Section 6.3 stated that the activated charcoal adsorber has a minimum iodine removal efficiency of 99 percent. In Section 10, Accident Analysis, SBGT filter efficiency was assumed to be 90% for halogen removal for the elemental and particulate forms of lodine, and 70% for the organic forms of lodine. Review of the Environmental Reports The initial PASNY Environmental Statement was submitted in May 1971. It did not contain discussions of accident environmental consequences, because it was not a requirement at that time. PASNY submitted Supplement 1 to the Environmental Report in November of 1971. Appendix L of this Supplement contained a discussion of accident environmental consequences as mandated by the newly revised 10CFR50 Appendix D (later 10CFR Part 51). App'andix L listed the LOCA inside the Drywell as a Class 8 accident. Section 12.1.1 of the Appendix states that the primary containment leakage rate used in the environmental analysis was 0.5 percent per day initial, with average 30 day release rate of 0.2 percent per day. Standby Gas Treatment filter efficiency was assumed to be 99.9% for removal of iodine. The analysis results show that doses resulting from a LOCA are orders of magnitude less than the limits stated in 10CFR20, and concluded that the environmental dose effects of this accident were of no significance. The assumptions used in Section 12.1.1 of Appendix L were consistent with the guidance that existed at the time in the form of a proposed Annex to Appendix D of 10CFR50 (36FR22851), the
" Draft Guide to the Preparation of Environmental Reports for Nuclear Power Plants," published by the AEC, dated February 1971, and the Supplement to the Draft Guide dated September 1971.
This fact was acknowledged in the NRC Final Environmental Statement issued in March 1973. The Annex required that certain assumptions be made in discussion of accidents in the environmental report submitted under Appendix D to 10 CFR 50. For Class 8 accidents, the Annex calls for use of " realistic" building leakage rates as a function of time. and states:
"The highly conservative assumptions and calculations used in AEC safety evaluations are not suitable for environmental risk evaluation, because their use would result in a substantial overestimate of the environmental risk. For this reason, Class 8 events shall be evaluated realistically. Consequences predicted this way will be far less severe than those given for the same event in the safety analysis reports where more conservative evaluations are used."
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Attachment IV to JPN-98-007 HISTORY AND BACKGROUND INFORMATION PERTAINING TO THE DISCREPANCY BETWEEN THE FSAR AND NRC SAFETY EVALUATION REPORT ALLOWED LEAKAGE RATES The position regarding the use of " realistic" assumptions was supported by the NRC Final Environmental Statement by reference to the Annex, and by discussions regarding the same issue prompted by a question from the State. NYS stated that the Table (in the Environmental Report) shows an estimated dose for a large break LOCA of < 0.5 mrem. The FitzPatrick FSAR shows a 2-hour off-site dose of 970 mrem, with thyroid dose of 11,400 mrem. The State requested that the discrepancy be clarified. The response of PASNY, and acknowledged by the NRC, was that the assumptions used in the FSAR calculations are different from those used in the Environmental Statement, and the results express upper limit vs. expected effects of such accidents. The conclusion that can be drawn from reviewing the Environmental Reports and the regulations that existed at the time of FitzPatrick licensing, is that the assumptions used in the Environmental Reports and those used in the FSAR bore no relationship to each other. The assumption for containment accident leakage provided by PASNY at the time was a realistic one based on design leakage of the containment (i.e. ty). The assumptions used in the accident analysis were based on worst case number which represented the maximum allowable containment leakage rate of 1.5 percent per day (i.e. L.). Therefore, amending the TS to clarify the difference between design and allowable leakage has no impact on the environmental statement. l I l l l l Page 6 of 6 l}}