ML20108D117
| ML20108D117 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 04/24/1996 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20108D109 | List: |
| References | |
| NUDOCS 9605070282 | |
| Download: ML20108D117 (68) | |
Text
Attachment I to JPN-96-016 PROPOSED TECHNICAL SPECIFICATION CHANGES (JPTS-96-003) f
}
l, New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 l DPR-59 l
1 l
9605070282 960424 PDR ADOCK 05000333 P PDR 1
JAFNPP TABLE OF CONTENTS (Cont'd) 6.16 Process Control Program (PCP) 258b 6.17 Offsite Dose Calculation Manual (ODCM) 258b 6.18 Major Modifications to Radioactive Liquid, Gaseous, and 258c Solid Waste Treatment Systems 6.19 Postaccident Sampling Program 258e !
6.20 Primary Containment Leakage Rate Testing Program 258e I ;
i 7.0 References 285 I I
I 1
1 I
l 1
Amendment No. 7,22,93,130, iv
l JAFNPP .
l LIST OF TABLES i
Table _ Tit!A Paae l 4.2-8 Minimum Test and Calibration Frequency for Accident Monitoring 86 '
l Instrumentation l
4.6-1 Snubber Visual inspection Interval 161 4.6-2 Minimum Test and Calibration Frequency for Drywell Continuous 162a Atmosphere Radioactivity Monitoring System
- 4.7-1 (DELETED) 210 4.7-2 (DELETED) 211 1 3.12-1 (DELETED) 244a 3.12-2 (DELETED) 244a 3.12-3 (DELETED) 244a 4.12 1 (DELETED) 244a 1
4.12 2 (DELETED) 244a l
l-4.12-3 (DELETED) 244a l l
6.2-1 Minimum Shift Manning Requirements 260a j 6.10-1 Component Cyclic or Transient Limits 261 !
l l
1 l
l t
l l
I 4
Amendment No. 20,22,92,ill,130,134,150,173,180,101,210,218, vi
. .- .. . - . . .-- - - - - _ . - - - - ~ - .. - ...- ~~- - .
l .r I f L t t
l f
.JAFNPP l i 4.0 BASES '
A. This specification provides that surveillance activities requirement will be identified as an exception. An example necessary to insure the Limiting Conditions for Operation are .
of an exception when the test interval is not specified in the met and will be performed during the OPERATIONAL regulations is the Note in Specification 6.20 " Primary CONDITIONS (modes) for which the Limiting Conditions for Containment Leakage Rate Testing Program," which states i Operation are applicable. Provisions for additional "The provisions of Specification 4.0.B do not epply to the surveillance activities to bo performed without regard to the test frequencies specified in the Primary Containment '
t applicable OPERATIONAL CONDITIONS (modes) are provided Leakage Rate Testing Program." This exception is provided in the individual Surveillance Requirements. because the program already includes provisions for i
extension of intervals. f B. Specification 4.0.B establishes the limit for which the specified time interval for Surveillance Requirements may be C. This specification establishes the failure to perform a
, extended. It permits an allowable extension of the normal Surveillance Requirement within the allowed surveillance ;
surveillance interval to facilitate surveillance scheduling and interval, defined by the provisions of Specification 4.0.B. as ;
consideration of plant operating conditions that may not be a condition that constitutes a failure to meet the l suitable for conducting the surveillance (e.g., transient OPERABILITY requirements for a Limiting Condition for '
conditions or other ongoing surveillance or maintenance activities). It also provides flexibility to accommodate the Operation. Under the provisions of this specification, systems and components are assumed to be OPERABLE l
length of a fuel cycle for surveillances that are performed at -[
each refueling outage and are specified with a 24 month when Surveillance Requirements have been satisfactorily performed within the specified time interval. However, surveillance interval. It is not intended that this provision be nothing in this provision is to be construed as implying that used repeatedly as a convenience to extend surveillance systems or components are OPERABLE when they are found I intervals beyond that specified for surveillances that are not performed during refueling outages. The limitation of this or known to be inoperable although still meeting the Surveillance Requirements. This specification also clarifies specification is based on angineering judgement and the that the ACTION requirements are applicable when recognition that the most probable result of any particular Surveillance Requirements have not been completed withm surveillance being performed is the verification of conformance with the Surveillance Requirements. The limit the allowed surveillance interval and that the time limits of the ACTION requirements apply from the point in time it is on extension of the normal surveillance interval ensures that identified that a surveillance has not been performed and not .I the reliability confirmed by surveillance activities is not significantly reduced below that obtained from the specified at the time that the allowed surveillance was exceeded. i Completion of the Surveillance Requirement within the surveillance interval. The exceptions to Specification 4.0.B allowable outage time limits of the ACTION requirements '
are those surveillances for which the 25% extension of the restores compliance with the requirements of Specification interval specified does not apply. .These exceptions are 4.0.C. However, this does not negate the fact that the stated in the individual Technical Specifications. The failure to have performed the surveillance within the allowed i
requirements of regulations take precedence over the survaillance interval, defined by the provisions of
'i Technical Specifications. Therefore, when a test interval is Specification 4.0.B, was a violation of the OPERABILITY specified in the regulations, the test interval cannot be requirements of a Limiting Condition for Operation that is extended under the provisions of 4.0.B. and the surveillance subject to enforcement ection. Further, the failure to 1
Amendment No. 83.188,19S,227, 30e l
JAFNPP 4.0 BASES - Continued C. Continued C. Continued 1 perform a surveillance within the provisions of Specification Surveillance Requirements do not have to be performed on 4.0.B is a violation of a Technical Specification requirement inoperable equipment because the ACTION requirements and is, therefore, a reportable event under the requirements define the remedial measures that apply. However, the of 10 CFR 50.73(a)(2)(i)(B) because it is a condition Surveillance Requirements have to be met to demonstrate prohibited by the plant Technical Specifications.' that inoperable equipment has been restored tc OPERABLE status.
If the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or a shutdown is required to comply with ACTION requirements, a 24-hour D. This specification establishes the requirement that all allowance is provided to permit a delay in implementing the applicable surveillances must be met before entry into an ACTION requirements. This provides an adequate time limit OPERATIONAL CONDITION or other condition of operation to complete Surveillance Requirements that have not been specified in the Applicability statement. The purpose of this performed. The purpose of this allowance is to permit the specification is to ensure that system and component completion of a surveillance before a shutdown is required to OPERABILITY requirements or parameter limits are met comply with ACTION requirements or before other remedial before entry into an OPERATIONAL CONDITION or other measures would be required that may preclude completion of specified condition associated with plant shutdown as well a surveillance. The basis for this allowance includes as startup.
consideration for plant conditions, adequate planning, availability of personnel, the time required to perform the Under the provisions of this specification, the applicable surveillance and the safety significance of the delay in Surveillance Requirements must be performed within the completing the required surveillance. This provision also specified surveillance interval to ensure that the Limiting provides a time limit for the completion of Surveillance -
Conditions for Operation are met during initial plant startup Requirements that become applicable as a consequence of or following a plant outage.
OPERATIONAL CONDITION (mode) changes imposed by ACTION requirements and for completing Surveillance When a shutdown is required to comply with ACTION Requirements that are applicable when an exception to the requirements, the provisions of this specification do not requirements of Specification 4.0.C is allowed. If a apply because this would delay placing the facility in a lower surveillance is not completed within the 24-hour allowance, CONDITION of operation.
the time limits of the ACTION requirements are applicable at that time. When a surveillance is performed within the 24-hour allowance and the Surveillance Requirements are not met, the time limits of the ACTION requirements are applicable at the time the surveillance is terminated.
Amendment No.19, Si, SS,109,162,183, 227, 30f
JAFNPP 3.7 (cont'd) 4.7 (cont'd)
(2) During testing which adds heat to the suppression pool, the water temperature shall not exceed 10'F above the normal power operation limit specified in ,
(1) above. In connection with such testing, the pool temperature must be reduced to below the normal l
power operation limit specified in (1) above within 24 i hours.
(3) The reactor shall be scrammed from any operating
- condition if the pool temperature reaches 110*F.
Power operation shall not be resumed until the pool temperature is reduced below the normal power operation limit specified in (1) above.
(4) During reactor isolation conditions, the reactor pressure vessel shall be depressurized to less than ,
200 psig at normal cooldown rates if the pool temperature reaches 120'F.
- 2. Primary containment integrity shall be maintained at all times 2. a. Perform required visual examination and leakage rate when the reactor is critical or when the reactor water testing of the Primary Containment in accordance temperature is above 212 F, and fuel is in the reactor with the Primary Containment Leakage Rate Testing vessel, except while performing low power physics tests at Program.
atmospheric pressure at power levels not to exceed 5 MWt.
- b. Demonstrate leakage rate through each MSIV is s 11.5 scfh when tested at a 25 psig. The testing ;
frequency is in accordance with the Primary '
Containment Leakage Rate Testing Program.
- c. Once per 24 months, demonstrate the leakage rate of 10AOV-68A,8 for the Low Pressure Coolant injection system and 14AOV-13A,B for the Core Spray system to be less than 11 scfm per valve when pneumatically tested at a 45 psig at ambient temperature, or less than 10 gpm per valve if hydrostatically tested at 2 1000 psig at ambient temperature.
Amendment No. 44, 166
JAFNPP Pages 167 through 175 Have Been Deleted Amendment No.
167 (Next page is 176)
JAFNPP 4.7 BASES (cont'd) assumption of no holdup in the secondary containment, 3. The MSIVs are tested at a pressure less than P, and resulting in a direct release of fission products from the ;t 25 psig, with a leakage rate acceptance criteria of primary containment through the filters and stack to the s 11.5 scfh per valve. This exemption was approved environs. Therefore, the specified primary containment leak by the NRC in the original Technical Specifications rate and filter efficiency are conservative and provide (Table 4.7-2).
additional margin between expected offsite doses and 10CFR100 guidelines. The Program as implemented meets the requirements of Option B of 10 CFR 50 Appendix J (16) and Regulatory The leakage rate testing program was originally based on Guide 1.163 (13), with the exception stated in Specification NRC guidelines for development of leak rate testing and 6.20. This exception applies to valves currently installed in surveillance schedules for reactor containment vessels. this configuration, and does not apply to new installations. t Containment structural integrity is currently verified with This exception is consistent with TS Table 4.7-2, previously visual inspections and containment leak tightness is verified contained in the TS, which allows reverse direction testing of by the leakage rate surveillance testing described in the valves as an exception to the requirements of the draft JAFNPP Primary Containment Leakage Rate Testing - Appendix J, on the basis that pressurization direction was ,
Program. not a requirement at the time of plant design.
The following are the exemptions to 10 CFR 50 Appendix J, Option A, that have been approved by the NRC, and remain applicable to Option B of 10 CFR 50, Appendix J:
1 The Type C exceptions listed on Table 4.7-2,
" Exception to Type C Test", as of the date of B. Slandby Gas Treatment System and -
issuance of Amendment 194 (July 29,1993). C. Secondarv Containment
- 2. Valves which are sealed with fluid from a seal Initiating reactor building isolation and operation of the system, such as the liquid in the suppression Standby Gas Treatment System to maintain at least a 1/4 in.
chamber are not required to be Type C tested. This of water vacuum within the secondary containment provides exemption was approved by the NRC in the original an adequate test of the operation of the reactor Technical Specifications (SR 4.7.A.2.c(3)).
Amendment No.97-134, 194
JAFNPP Pages 198 through 213 Have Been Deleted Amendment No. 48,91,'18,150,'?3, 198 (Next page is 214)
JAFNPP 6.19 POSTACCIDENT SAMPLING PROGRAM A program shall be established, implemented, and maintained which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, 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 leakage rate testing of the Primary Containment as required by 10 CFR 50.54 (o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with
- the guidelines contained in Regulatory Guide 1.163, " Performance-Based l Containment Leak-Test Program", dated September 1995, as modified by the exception that Type C testing of valves not isolable from the containment free air space may be accomplished by oressurization 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 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 I coolant accident (P ), is 45 psig.
B. The maximum allowable Primary Containment leakage rate (L ), at P., sha!! be 1.5% of primary containment air weight per day.
C. The leakage rate acceptance criteria are: ;
- 1. Primary containment leakage rate acceptance criteria is i 1.0 L,.
, During unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 10.60 L, for the Type B and Type C tests and 10.75 L, for the Type A tests:
- 2. Airlock testing acceptance criteria are:
- a. Overall airlock leakage rate is 10.05 L, when tested at A P ,
- b. For each door seal, leakage rate is i 120 scfd when tested at A P,.
D. The provisions of Specification 4.0.B do not apply to the test frequencies specified in the Primary Containment Leckage Rate Testing Program.
E. The provisions of Specification 4.0.C are applicable to the Primary Containment Leakage Rate Testing Program.
Amendmant No. 460, 258e
7.0 REFERENCES
(1) E. Janssen, " Multi-Rod Burnout at Low Pressure," ASME Paper (9) C.H. Robbins, " Tests of a Full Scale 1/48 Segment of the 62-HT-26, August 1962. Humbolt Bay Pressure Suppression Containment,"
GEAP-3596, November 17,1960.
(2) K.M. Backer, " Burnout Conditions for Flow of Boiling Water in -
Vertical Rod Clusters," AE-74 (Stockholm, Sweden), May (1C) " Nuclear Safety Program Annual Progress Report for Period 1962. Ending December 31,1966, Progress Report for Period Ending December 31,1966, OF;NL-4071."
(3) FSAR Section 11.2.2.
(11) Section 5.2 of the FSAR.
(4) FSAR Section 4.4.3.
(12) TID 20583, " Leakage Characteristics of Steel Containment (5) 1.M. Jacobs, " Reliability of Engineered Safety Features as a Vessel and the Analysis of Leakage Rate Determinations."
Function of Testing Frequency," Nuclear Safety, Vol. 9, No. 4, kly-August 1968, pp 310-312. (13) Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program", dated September 1995.
(6) Deleted (14) Section 14.6 of the FSAR.
(7) 1.M. Jacobs and P.W. Mariott, APED Guidelines for Determining Safe Test Intervals and Repair Times for (15) ASME Boiler and Pressure Vessel Code, Nuclear Vessels, Engineered Safeguards - April 1969.
Section Ill. Maximum allowable internal pressure is S2 psig.
(8) Bodega Bay Preliminary Hazards Report, Appendix 1, Docket (16) 10 CFR Part 50, Appendix J, " Primary Reactor Containment 50-205, December 28,1962.
Leakage Testing for Water-Cooled Power Reactors, Option B -
Performance Based Requirements", Effective Date October 26, 1995 (17) Deleted i
Amendment No. 190,227, 285
. _ _ . _ _ . _ _ _ _ _ . - - _ _ _ _ _ _ m._4__m-___m.ur.u.___ _ _ . _ _ _ -. ___m___ am_____.__ - _ _ _ _ _ _ _ _ _ _________ -___ _ _ _ .__ _ AGiu
Attachment ll to JPN-96-016 REVISED SAFETY EVALUATION FOR Pj40 POSED TECHNICAL SPECIFICATION CHANGES (JPTS-96-003)
(Changes are identified by revision bars located in the right margin) j l
l
)
i i
1 i
1 New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59
1 Attachment Il to JPN-96-016 SAFETY EVALUATION - REVISION 1 1 Page 1 of 18
- 1. DESCRIPTION OF THE PROPOSED CHANGES These proposed Technical Specification (TS) changes support adoption of the primary containment leakage rate testing requirements of Option B to 10 CFR 50, Appendix J (Option B), and clarify the numerical value of the allowable containment leakage rate (L ) as 1.5 percent per day. The specific changes are as follows:
- 1. Page iv, Table of Contents, add 6.20, " Primary Containment Leakage Rate Testing Program" and "page 258e" to the Table of Contents.
- 2. Page vi, List of Tables, denote that Table 4.7-2, " Exception to Type C Tests," is deleted. The revised text reads: i "4.7-2 (DELETED) 211"
- 3. Bases 4.0.B, page 30e, insert the following after the current discussion: ;
"The exceptions to Specification 4.0.B are those surveillances for which the 25%
extension of the interval specified does not apply. These exceptions are stated in the individual Technical Specifications. The requirements of regulations take precedence over the Technical Specifications. Therefore, when a test interval is specified in the regulations, the test interval cannot be extended under the provisions of 4.0.B. and the surveillance requirement will be identified as an exception. An example of an exception when the test interval is not specified in the regulations is the Note in !
Specification 6.20, " Primary Containment Leakage Rate Testing Program," which ;
states "The provisions of Specification 4.0.B do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program." This exception is provided because the program already includes provisions for extension of intervals."
Note: This change results in the last five lines of the second column being moved to j Page 30f.
- 4. Pages 166 through 174, delete the following SRs:
1 4.7.A.2.a (1) through (10),
4.7.A.2.b (1), (2) 4.7.A.2.c (1) through (5),
4.7.A.2.e (1) through (6),
4.7. A.2.f.
1 3
Attachment 11 to JPN-96-016 SAFETY EVALUATION - REVISION 1 Page 2 of 18
- 5. Page 166, insert the following new SRs:
"4.7.A.2.a Perform required visual examination and leakage rate testing of the Primary Containment in accordance with the Primary Containment Leakage Rate Testing Program."
4.7.A.2.b Demonstrate leakage rate through each MSIV is s 11.5 scfh when
- tested at 2 25 psig. The testing frequency is in accordance with the Primary Containment Leakage Rate Testing Program.
- 6. Page 167, add note stating that pages 167 through 175 have been deleted. Add note under the page number stating "(Next Page is 176)."
- 7. Page 172, relocate SR 4.7.A.2.d (1) to page 166 and renumber as SR 4.7.A.2.c. Make editorial changes to improve readability. The revised SR reads:
"4.7.A.2.c Once per 24 months, demonstrate the leakage rate of valves 10AOV-68A,8 for the Low Pressure Coolant injection system and 14AOV-13A,B for the Core Spray system to be less than 11 scfm per valve when pneumatically tested at 2 45 psig at ambient temperature, or less than 10 gpm per valve if hydrostatically tested at 21000 psig at ambient temperature."
i 8. Page 174, delete asterisked notes for one-time exemptions from the Type A,B and C testing requirements of 10 CFR 50 Appendix J.
- 9. Page 175, delete the intentionally blank page.
- 10. Bases 4.7.A, Page 194, delete the primary containment leakage rate testing discussion
- which starts on the first column, second paragraph. Replace with
"The leakage rate testing program was originally based on NRC guidelines for development of leak rate testing and surveillance schedules for reactor containment vessels. Containment structural integrity is currently verified with visual inspections and containment leak tightness is verified by the leakage rate surveillance testing l described in the JAFNPP Primary Containment Leakage Rate Testing Program.
a N
e
[ Attachment 11 to JPN-96-016
, SAFETY EVALUATION - REVISION 1 l Page 3 of 18 1
The following are the exemptions to 10 CFR 50 Appendix J, Option A, that have been approved by the NRC, and remain applicable to Option B of 10 CFR 50, Appendix J:
1
- 1. The Type C exceptions listed on Table 4.7-2, " Exception to Type C Tests," as of the date of issuance of Amendment 194 (July 29,1993).
- 2. Valves which are sealed with fluid from a seal system, such as the liquid in the ,
suppression chamber are not required to be Type C tested. This exemption was approved by the NRC in the original Technical Specifications (SR 4.7.A.2.c(3)).
- 3. The MSIVs are tested at a pressure less than P, and 2 25 psig, with a leakage rate acceptance criteria of s 11.5 scfh per valve. This exemption was approved -
by the NRO in the original Technical Specifications (Table 4.7-2).
i The Program as implemented meets the requirements of Option B of 10 CFR 50 Appendix J (16) and Regulatory Guide 1.163 (13), with the exception stated in Specification 6.20. This exception applies to valves currently installed in this configuration, and does not apply to new installations. This exception is consistent ,
with TS Table 4.7-2, previously contained in the TS, which allows reverse direction testing of valves as an exception to the requirements of the draft Appendix J, on the basis that pressurization direction was not a requirement at the time of plant design."
- 11. Page 198, Change page "209" to page "213," and change next page from "210" to "214" to reflect the deletion of page 211 through 213b discussed in item 13. .
- 12. Page 210, Delete the intentionally blank page.
- 13. Page 211 through 213b, Table 4.7-2 " Exception to Type C Tests," delete pages and ,
relocate Table in its entirety to the Containment Leakage Rate Testing Program.
1 t !
- m. _ _ .. . . . - _ . . _ . . _ _ _ . . ____ _ _ __
Attachment 11 to JPN-96-016 SAFETY EVALUATION - REVISION 1 Page 4 of 18
- 14. Page 258e, Administrative Controls, add new section 6.20 entitled " Primary Containment Leakage Rate Testing Program." The new section reads:
"6.20 Primarv Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the Primary Containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, 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 free 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 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 intemal pressure for the design basis loss of coolant accident (P,), is 45 psig. ,
l B. The maximum allowable Primary Containment leakage rate (L.), at P , shall be l 1.5% of primary containment air weight per day. !
C. The leakage rate acceptance criteria are:
- 1. Primary containment leakage rate acceptance criteria is s 1.0 L,.
During unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 L, for the Type B and Type C tests and s 0.75 L, for the Type A test;
- 2. Airlock testing acceptance criteria are:
- a. Overall airlock leakage rate is s 0.05 L, when tested at 2 P ,
- b. For each door seal, leakage rate is s 120 SCFD when pressurized to 2 P,.
Attachment 11 to JPN-96-016 SAFETY EVALUATION - REVISION 1 Page 5 of 18 D. The provisions of Specification 4.0.B do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.
E. The provisions of Specification 4.0.C are applicable to the Primary Containment Leakage Rate Testing Program."
- 15. Page 285, delete Reference 13 as it no longer applies under Option B, and replace with reference to Regulatory Guide 1.163. The revised Reference 13 reads:
"(13) Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995."
- 16. Page 285, revise Reference 16 to reflect new revisions to 10 CFR 50 Appendix J. The revised reference reads:
"(16) 10 CFR Part 50 Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Option B-Performance Based Requirements," Effective date October 26,1995.
- 17. Page 285, delete Reference 17 as it no longer applies under Option B, and replace with " Deleted."
ll. PURPOSE OF THE PROPOSED CHANDES The proposed changes to the TS support adoption of the primary containment leakage rate testing requirements of Option B at the FitzPatrick plant, and clarify the numerical value of the allowable containment leakage rate (L ) as 1.5% per day.
The original FitzPatrick TS were written prior to the effective date of the 10 CFR 50 Appendix J regulation. Therefore, to ensure adequate primary containment leak rate testing, many of the requirements of the draft Appendix J requirements were incorporated directly into the Specifications. Exceptions were included as part of the initial issuance of the TS. Consequently, primary containment leak rate testing requirements duplicate tho.se contained in the draft Appendix J (Option A), with approved exemptions.
To simplify implementeion of Option B, and prevent conflicts with the TS currently based on Option A, the Authority proposes to delete the following SRs:
4.7.A.2.a(1) through (10) Type A testing requirements - Option A Section Ill.A.
4.7.A.2.b(1) and (2) Type B testing requirements - Option A Section Ill.B.
4.7.A.2.c(1), (2) Type C testing requirements - Option A Section Ill.C
3
! Attachment il to JPN-96-016 SAFETY EVALUATION - REVISION 1 Page 6 of 18
, 4.7.A.2.c(3) Approved exemption that eliminates Type C testing for valves that are sealed from a seal system, such as the water in the suppression chamber. This exemption is l retained and relocated to the Program.
4.7. A.2.c.4 Reference to Table 4.7-2, " Exception to Type C tests."
The Table is relocated to the Program.
1 4.7.A.2.e(1), (2) Periodic retest schedule requirements - Option A Section Ill.D 4.7.A.2.e(3) through (4) Requirements for Type B tests of airlocks and testing of airiock seals. These SRs were exemptions from Option A testing requirements that are no longer required for implementation of. Option B. Acceptance criteria is relocated to proposed Specification 6.20.
4.7.A.2.e(5) Type C retest schedule - Option A Section Ill.D.
4.7.A.2.f Containment modification requirements - Option A Section IV.A The Authority proposes to relocate Table 4.7-2, " Exception to Type C Tests," from the ,
TS to the Program. This table lists the approved exemptions to Appendix J for Type C tests at the FitzPatrick Plant, and is still applicable to Option B.
The above SRs and Table are replaced by: (1) New SR 4.7.A.2 (a) which requires visual examination and leak rate testing of the Primary Containment in accordance with ,
the Program; and (2) New SR 4.7.A.2 (b) which incorporates an existing exemption for .
Type C testing of Main Steam Isolation Valves (MSIV) directly into the TS for !
consistency with References 2 and 3. A description of the Program, including plant I specific leakage limits, is contained in new Specification 6.20. Regulatory Guide 1.163 l (Reference 1) is incorporated, by reference, into new Specification 6.20 as required by Option B, Paragraph V.B.1.
Implementation of Option B requirements will be controlled under the Primary Containment Leakage Rate Testing Program (the Program), in accordance with the requirements of Option B, Regulatory Guide 1.163 (Reference 1), and exemptions from Option A currently approved by the NRC for the FitzPatrick plant. The proposed !
TS changes are consistent with guidelines provided in NUREG-1433 (Reference 2) I and NRC letter dated November 2,1995 (Reference 3), to the extent practicable.
There are differences between the FitzPatrick TS and Reference 2 relating to Containment Systems that will be resolved in our upcoming improved Standard Technical Specifications conversion effort. These differences do not adversely affect the TS changes required to support implementation of Option B at the FitzPatrick plant.
i
I 1
Attachment 11 to JPN-96-016 SAFETY EVALUATION - REVISION 1 Page 7 of 18 l
Ill. SAFETY IMPLICATIORS OF THE PROPOSED CHANGES l
This section discusses the safety implications of the TS changes relating to l implementation of Option B at the FitzPatrick Plant and the clarification of the L, i numerical definition. !
- 1. Option B Imolementation I The testhg requirements of 10 CFR 50, Appendix J, ensure that leakage through the l primary containment, including systems and components that penetrate the primary containment, does not exceed the allowable leakage rate values specified in the TS and bases. This ensures that an adequate primary containment boundary is maintained during and after an accident, thereby assuring that the primary containment function assumed u the safety analyses is maintained. l l
A revision to 10 CFR 50, Appendix J, to allow a performance-based approach to i containment leakage rate testing became effective on October 26,1995. The revision added Option B " Performance Based Requirements" to Appendix J to allow licensees to voluntarily replace the prescriptive testing requirements of Appendix J with testing requirements based on both overall and individual component leakage rate i performance. Option B allows plants with satisfactory Integrated Leak Rate Testing ;
(ILRT) performance history to reduce the Type A testing frequency from three tests in j ten years to one test in ten years. For Type B and Type C tests, the testing frequency 1 can be reduced based on the leak rate test history of each component. The Authority l has elected to perform Type A, Type B and Type C containment leak rate testing on a performance basis.
Regulatory Guide 1.163 (Reference 1) was issued by the NRC Staff as an acceptable method for implementing Option B. It states that NEl 94-01 (Reference 4) provides methods acceptable to the NRC staff for complying with Option B, with the four exceptions listed in Section C of the Regulatory Guide. The Authority will comply with the methods outlined in the Regulatory Guide, with the exception of Type C testing of I containment isolation valves in the reverse (non-accident) direction identified in proposed Specification 6.20.
The adoption of a performance-based primary containment leakage rate testing program does not change the method by which leakage rate testing is performed. The tests will continue to be performed at full pressure (P,) or greater, with the exception of existing NRC approved exemptions. Plant specific limits for allowable leakage rates (L ) and required test pressure (P,) are retained in the proposed TS and are not changed as a result of adopting Option B testing requirements. Testing methods will continue to be in accordance with existing leak rate testing requirements, as modified by exemptions previously approved by the NRC.
l l
l Attachment 11 to JPN-96-016 SAFETY EVALUATION - REVISION 1 Page 8 of 18 These changes do not alter the plant design, only the frequency of measuring primary containment leakage. Therefore, the proposed changes do not directly result in an increase in containment leakage. However, decreasing the test frequency can increase the probability that a large increase in containment leakage could go undetected for an extended period of time. NUREG 1493, " Performance- Based Containment Leak-Test Program, Final Report," (Reference 7) made the following observations with regard to the decreased test frequency:
= Reducing the Type A (ILRT) testing frequency from the current three per ten years to one per 20 years was found to lead to an imperceptible increase in ,
risk. The estimated increase in risk is small because ILRTs identify only a few I potential leakage paths that can not be identified by Type B and Type C testing, and the leaks that have been found by Type A tests have only been marginally above existing requirements. Given the insensitivity of risk to containment leakage rate, and the small fraction of leakage detected solely by Type A testing, increasing the interval between ILRT testing has minimal impact on public risk.
= While Type B and C tests identify the vast majority (greater than 95 percent) of all potential leakage paths, performance-based altematives to c"rrent local leakage testing requirements are feasible without significant risk impacts. The risk model used in NUREG-1493 suggests that the number of components tested would be reduced by about 60 percent with less than a three-fold increase in the incremental risk due to containment leakage. Since under l existing requirements leakage contributes less than 0.1 percent of overall accident risk, the overall impact is very small. i Option B states that specific exemptions to Option A of Appendix J, that have been formally approved by the NRC or AEC, are still applicable to Option B if necessary, unless specifically revoked by the NRC. The following exemptions to Option A will be retained in the Option B Program:
- 1. The Type C exceptions listed on Table 4.7-2, " Exception to Type C Tests," as of the date of issuance of Amendment 194 (July 29,1993).
- 2. Valves which are sealed with fluid from a seal system, such as the liquid in the suppression chamber are not required to be Type C tested. This exemption was approved by the NRC in the original Technical Specifications (SR 4.7. A.2.c(3)).
- 3. The MSIVs are tested at a pressure less than P, and a 25 psig, with a leakage l rate acceptance criteria of s 11.5 scfh. This exemption was approved by the l NRC in the original Technical Specifications (Table 4.7-2).
l Attachment 11 to JPN-96-016 SAFETY EVALUATION - REVISION 1 Page 9 of 18 These exemptions focus on the testing methodology aspects of Appendix J and are unaffected by the adoption of Option B testing frequency requirements. A list of approved exemptions will be contained in the Bases of the FitzPatrick TS, Section 4.7.A. The details of these exemptions will be contained in the Program.
Changes to the Program will be controlled in accordance with the requirements of 10 CFR 50.59, " Changes, tests and experiments." Thus, a determination of whether Program changes require prior NRC approval will be performed. In addition,10 CFR ;
50 Appendix J Option B requires Licensee compliance with containment leakage rate i testing requirements as stated in the regulation, and included by reference in the proposed TS. Changes to the Program that conflict with the requirements of Option B, or documents referenced in Specification 6.20, require prior NRC approval. The combination of the 10 CFR 50.59 change process and the NRC approval process ]
assure proper control of changes to the Primary Containment Leakage Rate Testing Program.
Current SR 4.7.A.2.d (1) requires pneumatic or hydrostatic leakage rate testing of the LPCI and Core Spray injection testable check valves. These valves are pressure isolation valves that separate the high pressure reactor coolant system from the low pressure LPCI and Core Spray systems. The leakage test required by SR 4.7.A.2.d (1) is not a requirement of 10 CFR 50 Appendix J, therefore, it is not relocated from the TS. Editorial changes to improved readability are made to the SR, and it is moved to page 166 and renumbered as SR 4.7.A.2.c.
Based on the above discussion, removal of the containment leakage rate testing details, except for plant specific limits, from the TS is acceptable. The proposed TS and the Program comply with Regulatory Guide 1.163 (Reference 1) requirements, with the exception of reverse direction Type C testing of valves described in Section Ill.3 of this safety evaluation, and contain sufficient controls to ensure that the primary l containment structural integrity is inspected and maintained, and that leakage is limited i to values assumed in the plant safety analyses. Required surveillances will continue to I be performed in accordance with TS, written procedures, and instructions auditable by the NRC. Primary containment leakage rate requirements continue to remain an integral part of FitzPatrick plant operation. The changes to current SR 4.7.A.2.d.(1) are l editorial in nature and do not change any TS requirement.
Attachment 11 to JPN-96-016 SAFETY EVALUATION - REVISION 1 Page 10 of 18
- 2. Clarification of the Numerical Definition of La Proposed Specification 6.20 defines the value of the allowable containment leakage rate (L,) as 1.5 percent of primary containment air weight per day. This is a clarification of the numerical value of L, for the FitzPatrick plant. This clarification was previously submitted as part of a proposed TS amendment (Reference 5), and subsequently withdrawn (Reference 6) 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 reactor building and has identified two component types that require further evaluation as a result of this clarification.
Appropriate action will be taken to qualify these component types prior to implementation of a 1.5 percent per day allowable leakage rate.
Current SR 4.7.A.2.a (8) defines the Type A acceptance criteria as less than 0.75 L, and not greater than the design leakage rate, Lo (0.5%/ day). This SR was contained in the original TS and was written to support the pre-operational test. The SR is consistent with the 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 Ill.A.5(b)(2), which defines the acceptance criteria as less than 0.75L,. For purposes of establishing Type A, B and C leakage test l acceptance criteria, the allowable containment leakage rate has been limited to 0.5 l weight percent of the contained air volume per day so as not to conflict with SR , 4.7.A.2.a (8). This interpretation is conservative with respect to the TS Bases, and the l current licensing basis. ;
4 This clarification potentially affects the off-site dose consequences of postulated l accidents which are directly related to containment leakage rate. The FitzPatrick accident analyses assumed an allowable leakage rate (L ) of 1.5 weight percent per day. The limitation on containment leakage rate ensures that total leakage 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, acceptance criteria stated in proposed Specification 6.20.
The effects of this clarification are: 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.
l i
Attachment ll to JPN-96-010 SAFETY EVALUATION - REVISION 1
- Page 11 of 18 l
- 3. Exception Regarding Reyftspe_Diffrlon Testing of 17 Primary Containment isolatiDD l ValyRS Periodic Type C testing in the reverse (non-accident) direction for 17 primary containment isolation valves does not expose potential atmospheric leakage paths (e.g., valve stem packing) to test pressure. Therefore, it can not be quantitatively shown that Type C test results are not affected in a non-consentative manner by directionality. Section 8.0 of Reference 4 requires that potential leakage paths to atmosphere be quantitatively determined. Reverse direction testing of these valves is required due the inability to isolate the valvea from the containment and the lack of test connections. These valves are reverse directon tested in accordance with the FitzPatrick TS Table 4.7-2, " Exception to Type G Tests."
The affected valves are listed in Table 1 of this safety evaluation. Type C testing in the reverse direction for these valves provides equivalent or more conservative results than testing in the accident direction, with respect to seat leakage. With respect to the
, globe valves, the test pressurization is under the seat, which tends to unseat the valve.
With respect to the butterfly valves, measured leakage is independent of the direction of test pressure from both a force exerted and seating surface standpoint.
Modifications have been considered that would allow testing in the accident direction or allow potential leakage to atmosphere to be quantitatively determined. The addition of block valves and test connections to allow accident direction testing would increase design complexity, provide additional potential leakage pathways, and increase loading on piping penetrating primary containment. Valve stem packing modifications to allow potential leakage to be quantitatively determined would increase design complexity, and provide additional potential leakage pathways. For these reasons, compliance with Reference 1 would incur an undue cost without a commensurate improvement in safety.
There are no safety implications associated with these changes because:
- 1. Testing of these 17 valves (Usted in Table 1) during the 1995 Integrated Leakage Rate Test (ILRT) verified that the packing glands were insignificant contributors to the overall integrated leakage rate. The 1995 as-left ILRT leakage rate was 0.0629% weight / day, which was well below the current TS acceptance criteria of 0.5% weight / day.
- 2. Adding the results of the 1995 As-Left Type A, B, and C tests together (approximately 2188 SCFD) results in a leakage total well below 0.6Le i (3216 SCFD). This very conservatively shows that significant margin exists to
! exceeding TS or Appendix J limits.
l t
Attachment 11 to JPN-96-016 SAFETY EVALUATION - REVISION 1 Page 12 of 18
- 3. Review of past ILRT results indicates that the 17 valves have not been the cause of an ILRT failure. Based on a review of the maintenance history for each valve, recurring packing or body to bonnet leaks are not expected.
- 4. The valve stem packing and body to bonnet gaskets are resilient materials designed to conform to sealing surfaces. The valves are installed in systems which are not normally subjected to design flows, temperatures, or pressures.
During normal operation, the valve stem packing and body to bonnet gaskets are exposed to the primary containment atmosphere, which has a low oxygen content. Based on this, the degradation of the valve stem packing or body to bonnet gaskets due to continuous exposure to a harsh environment is not a concem.
- 5. From a risk perspective evaluation, the elimination of modifications that would allow testing in the accident direction or allow potential leakage to atmosphere to be quantitatively determined, can be justified using the technical bases provided for NUREG-1493 (Reference 7). Past studies show that overall reactor accident risks are not sensitive to variations in containment leakage rate. This is because reactor accident risks a:; dominated by accident scenarios in which the containment fails or is bypassed. Such scenarios, even though 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 (See Table 2).
Certain NRC sponsored studies (References 8 and 9) indicate that overall plant l risk is not senstive to changes in containment leak rates. From Table 3 the incremental risk from leakage in the range of 1% to 10% per day is small.
FitzPatrick and Peach Bottom are both BWR 4 plants with MARK I containments. Similar results are expected for FitzPatrick.
The analysis described above provides justification that potential leakage paths to atmosphere for these 17 valves is inconsequential. The Authority proposes that a soap bubble test be performed on the pressurized stem / bonnet boundaries of the 17 valves during regularly scheduled Type A testing. To provide a direct indication of the leak-tightness of the packing and body to bonnet, the Authority will use the acceptance criteria of zero bubbles for this test. Type C testing will be performed, as a post work test, following work activities that affect the potential atmospheric leakage paths on any of the 17 valves. A soap bubble test will then be performed on the subject valve (s) at regularly scheduled Type A test intervals. These requirements will be contained in the Program.
IV. EVALUATION OF SIGNIFICANT HAZARDS CONSID_EBA'[10N
l l
Attachment il to JPN-96-016 SAFETY EVALUATION - REVISION 1 Page 13 of 18 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 10 CFR 50.92, operation of the James A. FitzPatrick Nuclear Power Plant in accordance with the proposed amendment will not:
- 1) Involve a significant inc.3ase in the probability or consequences of an accident previously evaluated becauce:
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 acc: dent previously evaluated is not increased by this proposed TS change. The clarification of the allowable containment leakage rate (L,) is consistent with the accident analyses.
There is no change to the consequences of an accident previously evaluated because maintaining leakage within limits assumed in the accident analyses ensures that the dose consequences resulting from an accident are not increased. The proposed TS changes maintain an equivalent level of reliability and availability for all affected systems. The ability of the affected systems associated with maintaining leak rate j integrity 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 within acceptance limits, !
and as such, will not significantly increase the consequences of a previously evaluated l accident.
- 2) Create the possibility of a new or different kind of accident from any accident previously evaluated because: j The proposed changes allow adoption of those requirements specified in Option B to 10 CFR 50, Appendix J, and do not involve a change to the plant design and operation. As a result, the proposed changes do not affect the parameters or conditions that could contribute to the initiation of any accidents. The methods of performing primary containment leakage rate testing are not changed. No new accident modes are created by allowing extended intervals for Type A, B and C testing, or by clarifying the numerical value of the allowable containment leakage rate (L ). 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 safety analyses.
Extension of the test intervals, and clarification of the allowable leakage rate, does not contribute to the possibility of a new or different kind of accident or malfunction from those previously analyzed.
l 3) Involve a significant reduction in the margin of safety because:
- - .. .. - - . -_ -- - _- = _
i Attachment ll to JPN-96-016 SAFETY EVALUATION - REVISION 1 Page 14 of 18 The proposed changes affect the frequency of primary containment leakage rate testing, and the numerical definition of the allowable containment leakage rate (L ).
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 a P,. The proposed changes provide sufficient controls to ensure that proper maintenance and repairs are performed on the primary containment, and systems and components penetrating the primary containment. The reliability of containment l systems assumed to operate in the plant safety analyses is not reduced. The l numerical value of L, specified in Specification 6.20 is consistent with the accident analyses, therefore, the dose consequences of any analyzed accidents are not '
increased. Therefore, the proposed changes provide continued assurance of the leak tightness of the containment without adversely affecting the public health and safety and, as such, will not involve a significant reduction in the margin of safety.
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 l settings or a significant relaxation of the bases for the limiting conditions for operations.
Therefore, based on the criteria established in 10 CFR 50.92(c), the proposed change does not constitute a significant hazards consideration.
1 l V. IMPLEMENTATION OF THE PROPOSED CHANGES 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.
The Authority requests NRC approval of this proposed amendment prior to July 1,1996 in order to adopt these changes prior to the upcoming Fall 1996 Refueling Outage.
VI. CONCLUSION Based on the discussions above, the adoption of Option B to 10 CFR 50, Appendix J, requirements into the TS will not decrease the effectiveness of containment leakage rate I testing. Operating limitations will continue to be imposed, and required surveillances will l 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 changes.
l The Plant 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.
k Attachment 11 to JPN-96-016 SAFETY EVALUATION - REVISION 1 2 Page 15 of 18 1
Vll. REFERENCES
- 1. Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995.
- 2. NUREG-1433, Standard Technical Specifications, Revision 1, dated April 1995.
- 3. NRC Letter to Mr. David J. Modeen (NEI), Regarding the Industry's Proposed Technical Specifications for Implementing Option B of Appendix J, dated November 2,1995.
- 4. NEl 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50 Appendix J, Rev. O, dated July 26,1995.
- 5. 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.
i
- 6. NYPA letter, R.E. Beedle to the NRC (JPN-92-016), " Withdrawal of Amendment Application (JPTS-84-012)," dated March 31,1992.
- 7. NUREG-1493, Performance-Based Containment Leak-Test Program, Final Report, i dated September 1995 1
- 8. 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 Desiga Safety Reviews," Volume 2, dated June 1,1986.
- 9. NUREG-1150, " Severe Accident Risks: An Assessment for Five U.S. Nuclear a j Power Plants, Final Summary Report," dated December 1,1990.
2 I
}
4 I
Attachment il to JPN-96-016 l SAFETY EVALUATION - REVISION 1 Page 16 of 18 l
l Table 1 i Additional Information on Valves l 1. Valve Number and
Title:
27AOV-112 - DRYWELL PURGE AND INERT ISOLATION VALVE
Description:
24", BUTTERFLY VALVE Vendor: FISHER CONTROLS CO. ,
- 2. Valve Number and
Title:
27AOV-113 - DRYWELL VENT AND PURGE EXHAUST INNER ISOLATION VALVE
Description:
24", BUTTERFLY VALVE Vendor: FISHER CONTROLS CO.
- 3. Valve Number and
Title:
27AOV-101 A - TORUS VACUUM BREAKER VB-6 ISOLATION VALVE
Description:
20", BUTTERFLY VALVE Vendor: FISHER CONTROLS CO.
- 4. Valve Number and
Title:
27AOV-101B - TORUS VACUUM BREAKER ~
VB-7 ISOLATION VALVE
Description:
20", BUTTERFLY VALVE Vendor. FISHER CONTROLS CO.
- 5. Valve Number and
Title:
27AOV-117 - TORUS EXHAUST INNER ISOLATION VALVE
Description:
20", BUTTERFLY VALVE Vendor: FISHER CONTROLS CO.
- 6. Valve Number and
Title:
. 27MOV-117 - TORUS VENT AND PURGE EXHAUST ISOLATION VALVES (27AOV-117 AND 27AOV-118)
INNER BYPASS VALVE
Description:
3", BUTTERFLY VALVE Vendor: FISHER CONTROLS CO.
- 7. Valve Number and
Title:
27AOV-116 - TORUS PURGE AND INERT ISOLATION VALVE
Description:
20", BUTTERFLY VALVE Vendor: FISHER CONTROLS CO.
- 8. Valve Number and
Title:
27AOV-131 A - CAD TRAIN A NITROGEN MAKE-UP ISOLATION VALVE
Description:
1.5", GLOBE VALVE Vendor: MASONEILAN INTERNATIONAL INC.
l I
Attachment Il to JPN-96-016
- SAFETY EVALUATION - REVISION 1 Page 17 of 18
~
- 9. Valve Number and
Title:
27AOV-131B - CAD TRAIN B NITROGEN MAKE-UP ISOLATION VALVE
Description:
1.5", GLOBE VALVE Vendor: MASONEILAN INTERNATIONAL INC.
4
- 10. Valve Number and
Title:
10MOV-31 A - RHR A CONTAINMENT SPRAY INBOARD ISOLATION VALVE
Description:
10", GLOBE VALVE Vendor: ANCHOR-DARLING IND.
- 11. Valve Number and
Title:
10MOV-31B - RHR B CONTAINMENT SPRAY INBOARD ISOLATION VALVE
Description:
10', GLOBE VALVE Vendor: ANCHOR-DARLING IND.
- 12. Valve Number and
Title:
10MOV-38A - RHR A TO TORUS SPRAY ISOLATION VALVE
Description:
4", GLOBE VALVE Vendor: WILLIAM POWELL CO.
i
- 13. Valve Number and
Title:
10MOV-38B - RHR B TO TORUS SPRAY ISOLATION VALVE
Description:
4", GLOBE VALVE Vendor: WILLIAM POWELL CO.
- 14. Valve Number and
Title:
27AOV-132A - CAD TRAIN A TORUS NITROGEN MAKE-UP ISOLATION VALVE
Description:
1.5", GLOBE VALVE Vendor: MASONEILAN INTERNATIONAL INC.
- 15. Valve Number and
Title:
27AOV-1328 - CAD TRAIN B TORUS NITROGEN MAKE-UP ISOLATION VALVE
Description:
1.5", GLOBE VALVE Vendor: MASONEILAN INTERNATIONAL INC.
i
- 16. Valve Number and
Title:
16-1 AOV-101 A - DRYWELL PRESSURE SENSING
Description:
3/8", PLUG TYPE GLOBE VALVE Vendor: COPES-VULCAN INC.
- 17. Valve Number and
Title:
16-1 AOV-102B - TORUS PRESSURE SENSING
Description:
3/8", PLUG TYPE GLOBE VALVE ,
Vendor: COPES-VULCAN INC. l l
Attachment Il tc JPN-96-016 SAFETY EVALUATION - REVISION 1 Page 18 of 18 Table 2 Conditional Containment Failure Probability Given Core Damage' Time / Containment Location Conditional Probability Early/Drywell Failure ,
0.536 Early/Wetwell Failure 0.068 Late /Drywell Failure 0.116
- Late /Wetwell Failure 0.144 No Failure 0.136 Table 3 Post Core Damage (Level 3) Comparison of Results Population Dose, person-rem / reactor year Peach Bottom Grand Gulf i
- Leak Rate % / day NUREG/ NUREG- NUREG/ NUREG CR- 1150 CR-4330 1150 4330 2 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 a
2 Containment venting considered as failure.
2 See Referenec R 3
See Reference 9 4
W
1 Attachment lll to JPN-96-016 MARKUP OF TECHNICAL SPECIFICATION PAGES (JPTS-96-003) l l
l l
l I
l New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59
.__. - - - _ - _ . . . _ . _=. . _ -
i i
1 1
JAFNPP i
Tiarr 0F CONTENTS (cont'd)
I PJEt 6.16 Process Control Program (PCF) 258b 6.17 Offsite Dose Calculation Manual (CDCM) 258b 6.18 Major Modifications to Radioactive Liquid, a Gaseous, and Solid Maste Treatment Systems 258c
} 6.19 Postaccident Sampling Program 258e
- o2. R.f.r..e.s 285 T
(jfgje h L4 lAl Y Amendmentso,((,f
JAFNPP LIST OF TABLES linblA IlllH Eage 4.2-8 Minimum Test and Ca:ibration Frequency for Accident Monitonn9 86 Instrumentation 4.6 1 Snubber Visual inspection interval 161 4.6 2 Minimum Test and Calibration Frequency for Drywell Continuous 162a Atmosphere Radioactivity Monitoring System 4.7-1 (DELETED) 210
/
4.7-2 [ Exception to Type C TW ( Dfr LE TED) 211 A
3.12-1 (DELETED) 244a 3.12 2 (DELETED) 244at 3.12-3 (DELETED) 244a{
4.12-1 (DELETED) 244a -
4.12 2 244a l
(DELETED) 4.12-3 (DELETED) 244a (
6.2 1 Minimum SNft Manning Requirements 260a 6.10-1 Component Cyclic or Transient Limits 261 Amendment No. 20,22,02,iii,100,104,150,15,100,101,210, N vi
JAFNPP 4.0 BASES
, A. This specification provides that surveillance activities C. Continued l necessary to insure the Limitmg Conditions for Operation are met and will be perforrned during the OPERATIONAL interval, defined by the provisions of Specification 4.0.B, as a CONDITIONS (modes) for which the Luniting Conditions for condition that constitutes a failure to meet the OPERABIL!TY Operation are appbcable. Provisions for additional survesliance requirements for a Limiting Condition for Operation. Under the activities to be performed without regard to the appiicable provisions of this specification, systems and components are OPERATIONAL CONDITIONS isnodes) are provided in the assumed to be OPERABLE when Survedlance Requirements endeveduel Survedience Requirements. have been satisfactorily performed within the specified time interval. However, nothmg in this provision is to be construed D. Specification 4.0.8 estabbahes the limit for which the specified as implying that systems or components are OPERABLE when time interval for Survesilance Requwements may be extended. they are found or known to be inoperable although still it permits an allowable extension of the normal survedlance meeting the Survedlance Requirements. This specification also interval to facilitate survedlance scheduhng and consideration clarifies that the ACTION requirements are applicable when of plant operating conditions that may not be suitable for Survedance Requirements have not been completed within the conducting the survedence (e.g., transient conditions or other allowed survesitance interval and that the time limits of the ongoing survedence or maentenance actwettes). It also
- ACTION requwements apply from the poent in time it is provides flexibility to accommodate the length of a fuel cycle identified that a surveellence has not been performed and not for surveellences that are performed at each refueling outage at the time that the allowed surveellence was' exceeded.
and are specified with a 24 month survedance interval. It is Completion of the Survesence Requwement within the not intended that this provisson be used repeatedly as a allowable outage time limits of the ACTION requirements convensence to extend survesilence intervals beyond that restores compliance with the requerements of Specification specified for survedences that are not performed dunne 4.0.C. However, this does not negate the fact that the failure
! refueling outages. The lunitation of this specification is based to have performed the survedance within the allowed on engensonne ==twa and the recognition that the most survedance interval, defmed by the provisions of Spec 3ecation probable result of any particular survedlance being performed 4.0.8, was a violation of the OPERABILITY requirements of a is the verification of conformance with the Survedance Limiting Condition for Operation that is subject to enforcement Requerements. The limit on extension of the normel action. Further, the failure _tglperform a surveillance within it:e }
surveillance interval ensures that the reliability confirmed by provisions of Specification 4.0.8 is a violation of a Technical survedence activities is not significantly reduced below that Specification requirement and is, therefore, a reportable event obtaened from the specified surveellance interval. < ] under the requirements of 10 CFR 50.73(a)(2)(il(B) because it is a condition prohibited by the plant Technical Specifications.
C. This specification establishes the failure to perform a l
Surveillance Requirement within the allowed surveillance k l g 4 3 moved b ed- fp Amendment No. S3,1 SS,1SS,227 -
! 3%
i i i
l t
INSERT A:
2 "The exceptions to Specification 4.0.B are those surveillances for which the 25% extension of the interval specified does not apply. These exceptions are stated in the individual Tect.nical j
Specifications. The requirements of regulations take precedence over the Technical Specifications. Therefore, when a test interval is specified in the regulations, the test interval
- cannot be extended under the provisions of 4.0.B. and the surveillance requirement will be 1
identified as an exception. An example of an exception when the test interval is not specified ,
l in the regulations is the Note in Specification 6.20, " Primary Containment Leakage Rate Testing Program", which states "The provisions of Specification 4.0.B do not apply to the test ,
i frequencies specified in the Primary Containment Leakage Rate Testing Program." This l exception is provided because the program already includes provlsions for extension of l Intervals."
i ,
l l
l l
d i
i 1
kom pW5 P*1'- JAFNPP k
4.0 BASES - Continued C. Continued C. Continued if the allowalde outage time Nedts of the ACTION requirements Survedlance Requirements do not have to be performed on are less then 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or a shutdown is required to comply inoperstdo equipment because the ACTION requwements
, with ACTION requirements, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance is provided to define the remedeel measures that apply. However, the permit a deley in implementing the ACTION requwements. Survedience Requirernents have to be met to demonstrate that This provides an adequeen time Emit to complete Survedisnce enoperable equipment has been restored to OPERABLE status.
Requwements that have not been performed. The purpose of this allowance is to pennit the completion of a survedlance before a shutdown is requwed to c-r.#; with ACTION D. This specification estabhshes the requirement that all requirements or before other remedeel measures would be apphcable survediences must be met before entry into an required that may preclude completion of a survedience. The OPERATIONAL CONDITION or other condition of operation basis for this aNowance ecludes consideration for plant specified in the Applicability statement. The purpose of this conditions, adequate plannmg, avedobitty of personnel, the specification is to ensure that system and component tima required to perform the survedience and the safety. OPERASILITY requwements or parameter limits are met before segnificance of the delay in completing the requisad entr ;into an OPERATIONAL CONDITION or other specified survestance. This provision also provides a time limit for the conJenen associated with plant shutdown as stell as startup.
completion of Surveellence Requwements that become apphcatdo as a consequence of OPERATIONAL CONDITION Unc'er the provisions of this specification, the applicable (mode) cF ages imposed by ACTION requirements and for Susvedlance P.equirements must be performed within the completing Survoegance Requerements that are applecable when specified survedlance interval to ensure that the Limiting an exception to the requirements of Specification 4.0.C is Conditions for Operation are met during initial plant startup or allowed. If a surveellence is not completed within the 24-hour following a plant outage.
allowance, the time limits of the ACTION requkements are opphcable at that time. When a survedience ,% performed When a shutdown is required to comply with ACTION within the 24-hour allowance and the Survedlance requwements, the provisions of this specification do not apply Requerements are not met, the time limits of the ACTION because this would delay placing the facility in a lower requwements are applicable at the time the surveenance is CONDITION of operation.
terminated.
Amendment No.10, Si, "",1^^,1S2,1SL 227-30f
3.7 (cor.t'd) .JAFNPP /,,7 (cont'd) d
) (2) During testing which adds heat to :he suppressioe' pool, the wct:r terperatuic shall not exceed ICF above the nomal power operation linit specified in (1) above. In connection with such testing, the pool ter;ieraturc z.ust be restuced to belaw ti.c norral power operation l ir. i t specified in (1) above within F b I 2-. hours.
(3) Tlic reactor shall be scrammed from any operatin;; condition if- the pool ]
te.mperature reaches 110F. Power
- 2. The pri:nry con tc:.I=ent cpesatica shall not be resuaal until ,
integrity chall be den:cas t rat e.1 the pcol ter.pcrature is reduced below 03 f0110"S the nar. mal poucr operation limit .
t specified in (1) ahuve. .
- a. Type A Test (primary Centainment Ir.t eg rc.t :* 1 (4) D.n rin:: reacter isolation conditions, the reactcr pressure vessel sha,ll bo Leakage Rate Test) c'epressuri:cil to less than 200 psi;; inspection
- r. t nort_al couldoun rates if the pool (1.) Contai nmen t shall be performe.1 a:; a te: perature reaches 120F. the prerequinite to Primary containment integrity shall be maintained 1:er f or nance of Type A
- 2. the at all tines when the reactor is critical or whea t e s t r.. Durin.1 the rea: tor water tennerature is above 2120F, and pe rioil bet.ueen Ihe init.iation of the feel is in the reacto'r vessel, except while ,
pe r fo r.:i n;: io.cpc.cr physics tests at atuospheric coa t-p niaen t i n.: pert ion i
press :re at pc..cr levels not to exceed 5 Irdt. " "'I the perfcsn ance OL' the Type A trat, .ia repairs. or ail ju:; ton nt s shall be made.
X Amendment No. W 166
INSERT B:
" 4.7. A. 2. a Perform required visual examination and leakage rate testing of the Primary Containment in accordance with the Primary Containment j Leakage Rate Testing Program.7 4.7.A.2.b Demor. ate leakage rate through each MSIV is s 11.5 scfh when teste e.( _>_25 psig. The testing frequency is in accordance with the . j Primary Containment Leakage Rate Testing Program. i J
l 4.7.A.2.c Once per 24 months, demonstrate the leakage rate of 10AOV-68A,B for the Low Pressure Coolant Injection system and 14AOV-13A,B for the Core Spray system to be less than 11 scfm per valve when i pneumatically tested at 145 psig at ambient temperature, or less 1
than 10 gpm per valve if hydrostatically tested at 11000 psig at ambient temperature "
l l
)
l i
)
i i l
1 i
1 k
l 1
4 6
i
JAFNPP /. 7 (cont'd)
(2.) Closure of containment, isolation valves for the Type A test shall be accomplished by normal operation and without any preliminary exercising.
hq3 as IO "0 (3.) The containment test he b ggQ conditions stabilize for a period shall of about 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to the start of a leakage rate test.
(4.) Components to be tested as part of the containment shall be vented to the containment atmosphere.
(5.) Test methods are to comply with ANSI N45.4-1972 paragraph 5 and leak rate calculations will comply with the intent of ANSI N45.4-1972 paragraph 5.
The mass of air in the containment will be calculated hourly and the leak rate determined by a linear least sq uares fit to the mass or air as a function of time.
167 (Nar Pu3e &
l%) - -
JAFNPP 4.7 (cont'd)
(6.) The accuracy of the ;
Type A test shall in-verified by a supplemental test as described in Appendix C of ANSI N45.4-1972, or {
the metered addition of air into the containment after the end of the fgh Type A test. .
(7.) Test Pressure (a . ) An initial test shall be performed .
at a 23 psig (Pt, reduced pressure) which is greater than 0.50 Pa to measure a leakage rate Ltm.
f (b.) A second test shall be performed at 4% psig (Pa peak rassure) to measure i leakage rate Lam.
(c.) The leakage characteristics yielded by measurements Ltm and Iam shall establish the maximum allowable test leakage rate Lt of not more than La (Ltm/ Lam) . In the event Ltn/ Lam is greater than 0.7, Lt shall be specified as equal to La 168 yPt/Pa).
JAFNPP 4.7 (cont'd)
(8.) Acceptance Criteria
, Reduced pressure tests.
l g,14 (Pt) The leakage rate Ltm shall be less than 0.75 Lt.
O P Peak pressure test.
(Pa) The leakage rate Lam shall be less than .
0.75 (La) and not greater than IA, which is 0.5 weight percent of the contained air per 24 hr at the test pressure Po.
(9.) Periodic leakage rate test shall be performed at reduced pressure (Pt) or at peak pressure (Pa).
(10.) Additional requirements If any periodic Type A test fails to meet the applicable acceptance criteria the test schedule applicable to subsequent Type A tests will be reviewed and approved by the Cosianission.
If two consecutive periodic Type A tests fail to meet the acceptance criteria, a 169
.-~ - . . . ,.
t JAFNPP 4.7 (cont'd)
Type A test shall be performed at each plant shutdown for refueling or approximately every 18 months, whichever occurs first, until two consecutive Type A tests meet the acceptance criteria.
- b. Type B tests (Localleak rate testing of contamment penetrations) . ,
N-<-
$}u f (1.) All preoperational and periodic Type B tests shall be l
performed by local pneumatic pressurization of the contaenment penetrations, either individually or in groups. I, <
at a pressure not less than Pa, and the gas flow to /
maintain Pa shall be measured.
(2.) Acceptance criteria f
The comtuned leakage rate of all penetrations and valves -
subsect to Type B and C tests shall be less than O.60 La, with the exception of the valves sealed with fluid from a seal system.
4 i
\
Amendment No. 1 5,190
- i 170
JAFNPP 4.7 (cont *d)
- c. Type C tests
)Ek4LC__
- 7) g (1.) Type C tests shall be performed by j local pressurization. The pressure shall be applaed an the same darectaon as that when the valve would be required to perform its safety function, escept as listed in Table 4.7-2 unless it can be determined that l the results from the tests for a pressure applied in a different direction will provide equivalent or more conservative results. Each valve to be tested shall be closed by normal operation and without any preliminary esercising or adjustments.
(2.) Valves, unless pressurized with fluid from a seal system, shall be pressurized with air or nitrogen at a pressure of Pa, and the gas flow to maintain Pa shall be measured.
(3.) Valves, which are sealed with fluid from a seal system, such as the liquid in thie suppression chamber shall not be tes te d.
Amendment No. pd, 134
~
JAFNPP 4.7 (cont'd)
(4.) See table 4.7-2 for exceptions.
f (5.) Acceptance criterion - The combined leakage rate for all penetrations and
{ valves subject to type B and C tests shall be less than 0.60 La. Leakage from containment isolation valves that are sealed with !luid Irom a seal system may be excluded when determining the combined leakage rate provided that the installed isolation valve seal-water '
system fluid inventory is sufficient to assure the sealing function for at least 30 days.
bd. Other leak rate tests (1) The leakage rate for containment isola-tion valves 10-A0V-68A, 8 (penetration X-13A, B) for Low Pressure Coolant l
- Injection system and 14-A0V-13A, B !
'p ge- lbb (penetration X-16A, B) for Core Spray j System shall be less than 11 cubic feet g g per minute per valve (pneumatically T.
G s tj *7, A.J . c. tested at 45 psig with ambaent temper-ature) or 10 gallons per minute per valve (hydrostatically) tested at 1000 psig with ambient temperature.
M 134 Amendment No. y u_u
JA 4.7 (cont'd)
- e. Periodic retest schedule. 5 (1.) Type A test.
i After the preoperational leakage rate tests, a set of three Type A tests shall be perforned, at j 6 approximately equal intervals during each 10-year service period.
T.
I I
l Amendment 40 172a
..______m.m.______-_____ ..___-___ __ -.- __ _ ___ _ _- __ - _
/
JAFNP --
4.7 (cont'd)
[)pjbefE-o< 4L-The third test of each set shall be conducted when the plant is shutdown C for the 10-year plant inservice inspections.
Permissible periods for testing. The performance of Type A tests shall_be limited to periods when the plant facility is nonoperational and secured in the shutdown condition under the administrative control and in accordance with the plant safety procedures. .
(2) Type B tests, (except tests for air-locks), shall be performed during each reactor shutdown for refueling, or other convenient intervals, but in no case at intervals greater than 2 years.
(3) Type B tests of airlocks shall be conducted at an internal pressure of not less than 45 psig (Pa). The overall leakage rate for the airlock shall be less than or equal to 268 SCFD (0.05 La). Airlock tests shall be conducted:
a) Every six months.
b) Prior to restoration of contain-
' ment integrity, when maintenance has been performed on the airlock which could affect its sealing capability.
Amendment No. 97
- 173
sw
[ JAFNF N
./
y k 4.7 (cont'd)
[#( b c) Within three days of opening the airlock, when containment integrity is required and maintenance has been performed on the airlock which could affect its sealing capability.
(4) Airlock seals shall be tested at a pressure not less than 45 psig.
The seal leakage rate shall be less than or equal to 120 SCFD.
Airlock ducted:
seal tests shall be con-a) Prior to restoration of contain-ment integrity *. If maintenance which could affect sealing capa-bility was performed the entire airlock shall be tested as required by 4.7. A.2.e (3).
b) Within three days after opening the airlock, when containment integrity is required.
c) Once every three days, during periods of frequent openings when containment integrity is re-quired.
i Amendr'en t No. 97 173a
- Exemption to 10 CFR 50, Appendix J l
JAFNPP 4.7 (cont'd)
(5) Type C test. '
MA-c Type C tests shall be performed during each l ;
! reactor shutdown for refueling but in no case at intervals greater than two years.*
- l (6) Other leek rate tests specified in Section 4.7.d shall be performed during each reactor shutdown for refueling but in no case at intervals greater than two years.
- f. Containment modification
- Any major modification, replacement of a component which is part of the primary reactor containment boundary, or ressaling a seal-welded door, performed after the preoperational leakege '
' rate test shall be followed by either a Type A, Type i B, or Type C test, as applicable, for the area affacted by the modification. The measured leakage from this test shall be included in the test report.
The acceptance ct;teria es appropriate, shall be met.
Mmor modifications, replacements, or resealing of seal-welded doors, performed directly prior to the conduct of a scheduled Type A test do not require a separate test.
In accordance with an exemption from 10 CFR 50 Appendix J, a "
in accordance with an exemption from 10 CFR 50 Appendix J. -
Type A, B, or C test is not required for the replacement of piping the Type C test of the shutdown coolmg isolation valves and welds which constitute the Core Spray System rrurumum flow (10MOV-17 and 10MOV-18) may be deferred until refueling lines (3*-W23-152-7A, B) during the 1993 maintenance outage. outage Reload 11/ Cycle 12.
Amendment No. g,JVf, g5,1[4,1)f6,1[,1%, 208 174
_ _ _ - _ __ .____ -__-__ - - - - - - - - - - --- -- ---- - ^ - - - ~ ^ ~ ~ ~ ^ - - ~ ^ ^
t i
- F l
l l
l- t f
I s.
t
\ca
' 1 -
5
\
k
)h am
+
X.
O i
- c i
JAF rr )
4.7 BASES (coat'd) casumption of no holdup la the secondary contain- TAsmostleakageanddeteriorationofintegrityis '
meat, resulting in a direct release of fission expected to occur through penetrations, especially products from the primary costalement through the those with resilient seals, a periodic leak rate filters and stack to the environs. Therefore, test program of such penetrations is conducted at the specified primary cesteiament leak rate and the peak pressure of 45 pelg to insure not aaly filter offaciency are commervative and provide that the leakage remalas acceptably low but also addittomal margia between espected offsite doses that the soallag materials can withstand the and 10CFB100 guidelines. accident pressure. For airlock leak test, a seal test at the peak pressure could be substituted fThenazimumallowabletestleakrateatthepeak,' for the complete airlock test, if no malatemance pressure of 45 peig (Pa) is 0.5 weight percent work is done which could affect the sealing per day (Lam). The maalaus allowable test leak capability of the airlock.
rate at the reduced pressure of 23 pelg (P g)
' will be verified to be conservative by actual The leak rate testing program was originally ;
primary costelament leak rate measurements at based on Commission guidelines for development of I l both 45 psig and 23 peig upon completion of the leak rate testing and surveillance schedules for containment structure. reactor contaissent vessels (16), and discussed l
in Question 5.4 of the FSAR. Olith the esceptions To allow a margia for possible leakage deterior- listed in Table 4.7-2, the system comforms to the ction between latervals, the nazimum allowable latest Commission guidelines (17). The exceptions leak rate (Ltm), which will be met to remala on l' (
stated la Table 4.7-2 are necessary since the normal test schedule, is 0.75 Lt . In additional requirements were added after the ;
addition, it is latended to operate the primary system was designed.
containment structure at a slight positive pressure to continuously monitor primary costala- 3. Etandby Gas Treat-st System '
meat leakage. C.
t -_ _) Secondary Contaimmant Initiating reactor building isolation and opera-tion of the Standby Gas Treatment System to malatain at least a 1/4 in. of water vacuum ZAN C C. within the secondary contalement provides an adequate test of the operatloa of the reactor Amendment tio. .W 194 i
4 4
INSERT C:
3 "The leakage rate testing program was originally based on NRC guidelines for development of leak rate testing and surveillance schedules for reactor containment vessels.
- Containment structural integrity is currently verified with visual inspections and containment leak tightness is verified by the leakage rate surveillance testing described in the JAFNPP Primary Containment Leakage Rate Testing Program.
1 The following are the exemptions to 10 CFR 50 Appendix J, Option A, that have been j approved by the NRC, and remain applicaNe to Option B of 10 CFR 50, Appendix J
- 1. The Type C exceptions listed on Table 4.7-2, " Exception to Type C Tests," as of l
- the date of issuance of Amendment 194 (July 29,1993). !
i 2. Valves which are sealed with fluid from a seal system, such as the liquid in the suppression chamber are not required to be Type C tested. This exemption was
- approved by the NRC in the original Technical Specifications (SR 4.7.A.2.c(3)).
- 3. The MSIVs are tested at a pressure less than P, and .;>_25 psig, with a leakage rate acceptance criteria of .111.5 scfh per valve. This exemption was approved by the NRC in the original Technical Specifications (Table 4.7-2)."
The Program as implemented meets the requirements of Option B of 10 CFR 50 Appendix l J (16) and Regulatory Guide 1.163(13), with the exception stated in Specification 6.20.
i This exception applies to valves currently installed in this configuration, and does not apply to new installations. This exception is consistent with TS Table 4.7-2, previously contained in the TS, which allows reverse direction testing of valves as an exception to the requirements of the draft Appendix J, on the basis that pressurization direction was not a
- requirement at the time of plant design."
5 l i
l l
i I
1,.,- a - a ,n== -J- --o r - a a as- --1 - - ,a i
~
' ~
l 5
l-t 9
4 M
r b
I .
l$
l 2
Y Y
4 1
y _ w _
1 i
l M
f
(
. 2 O
f k
' b E
m_
N s I W .
4
%v s P
]
l el bea e s d l
wrear d etao r e .
suwb u t t
s sAo t e
e r .e C Pd eth e
. r p nue osnn T y
iai t
c er e i
emer r t n
n e
b t
o dee o o s n e agdo) sk i
t c
i t
c s
s n a
e e e v aeeig r t e
ir r c c D e l ws l d i
d o r
h c
E rdep e e p i A e nv5 s s y h G r R had2 e e r
b w O
t ne aa sr t v e v d e F ivo r e r e l v
R dlabH m e a nF e e e v E
P t
eviC sneS ht n
ht n
s e
r e
e T te oh i t i i r h S etan1 5 d d a s E bl oo1 e e s T Esdte t
s s n e :.
~
e e n E wie sm i t
e t u T
A R
shuS et i
b b e s a C.v S l av neb e 'e. e e d e a T K ve ere t
w w t s n S A s a E E L
dw a e wk r t ge e
v s
e v
t e
e i
s T L obieel l a l a b e v
bdps v v C A ne 5 'e e e t
o l
e E C i i ep2w l
s e s n w _
2 O h pf o l e e s _
, - P L Taov h h i h
, 7. Y T T w T 4T
, EO L 1
,A
. BT A
1 2
TN O
I t s
T n P e E m A8C0A8cD ABt e a0o0s8ee 21 3 2 A C r R 1 33 1 2 X
i qu E8 E 84s8a444 VVVVVVVV
- - - 1 1 1 1- 1 -
0 4
E e VA OOOOOOOO VVV VV s u
1 r
g AU LR AAAAAAAA OOO AAA OO A o V 99999999 777 At i r E n VN 22222222 77 a 7 i
t s
222 22 v 0 e
t C
- e p
y T
m o
r e A ,
f s r r e
U o n "e ,s r
a o/
gd e
gd
/o m a
b a
Fs A N m e p r n gr o a r n c
=
sO uI T S e
t Pu Ei ar d e
ua Pr$
l i r
s n
e i nP se p,
= r_ C eAng l eAeg m r r n
+
e N i m w (t o w (t o u eo v
r a - a yer rde yer rl b me aC -
._ PU t n r n
+
s DIrN DInN e Ti n
i o
s t
t e
e .
p .
n e
p N N g E OI T
8 8 4~
c6 2
h 00 T &
w a
N A R A I
A 8 _
s A T A8CD 5 a 6
o TE 7777 2 2 3 6 m e l h
e NN OE 3 .
T CP _
W7
.wmorr '
i . ASLE G.7-2 EXCEMION TO TYPE C TESTS i
CONTAINhENT VALVE PENETRATION PENETRATION PUNCTION NUhMER LOCAL LEAK RATE TEST PERFOflMED
) '
35C Trowereing incese Probe "C" 07EV-104C This volve is an :-fi cheer velve which cannot be Typo C tested.
35D TrevereinS ImCore Probe "S" 07EV-1048 This wafwe le en emplosive sheer wafve which cannot be Type C tested.
37A Centrol Red IJrtue 03SOV-120 378 WIN not be tested se lines (there are 137 lines, with 31 to 38 lines per !
tholour placent 03SOV-123 37C penetreeien, and each has the four indmated wolves) are sealed by 03AOV-128 process fluid. '
37D 03CRD 138 ,
3SA Central Red Delve 03SOV-121 388 WW not be tooted as lines (there are 137 lines, with 31 to 38 lines per '
tehove pissent 03SOV-122 38C penseresien, and each hoe the three indicated valves) are sealed by !
03ADV-127 process Auld.
300 :
j I
~
h .
39A fMI Cent. Sprey 10MOV-31A This weeve wW be tested in the roweroe direccon. I 398 , OHL Cent. Spray 10MOV-318 This valve wW be sosted in the roweres aAreccon.
f 45 Dryses Ptosewa Sensing 16-1 AOV-101 A This wafwe we he tested in the roweroe direccon. ,
50C Instnenenteelen - SensinS Verleue '
y% These inseumont root valves are tested during a Type A test.
Amendment No. , ,
, jl d, [1,194
[
--- . - - - - . . . - . . - . . . . . . . . . . - - . . . . - - . ~ . . . - . . . _ - - - . . . . . . ~ _ _ . . . _ . _ _ _ _ . . _ _
j N
1 I .I l i .
i i i i - 1 l
lr '
1 a i 1 l l
- I I 3 l l 1 1 I: i "
\
', i a l
! i '
1 1 i - 1 1 i > >
j b)ei m
e .
I I 1 li 11 a
~l" ii. as us!=n is a a
, E! !! ll !!!!!! !!n!!!!! !
! j fj fj i 1 l
5 I,
>1,8 nl. l.;3 .l.
RE; 3
. s 1& 3 152 i 11 %
i8 l d' I " ! ! 5 l gl l c4
I-
\q I
i l ,l I I j j j , ,
i, l I~ l 1 ,l l ,
l i l 3 3 ' y E
1 I
1 I
s
,! 1 1 ,I
- i. i.
i i
i i i ; ' 1 1 1 1 1 g ,
I ;,[ l I I i n
i n .
ge 1. s i 1. s s. .si si si 2
! -n an on l - ,-
il ! l ill i i i
!!!!i lli
- \ l f ,} i s j l
i
,l' !!
o 1 11 !
i s, i n ni I
n!!il 1
1 1 II e
!g Ig i a a a a i !
l av
~
JAFNPP N_
r T ABLE 4.7 ,,
EXCEPTION TO TYPE C TESTS CONTAINMENT PENETRATION VALVE PENETRATION LOCAL LEAK RATE TEST PERFORMED FUNCTION NUMBER
??G HPCI - Pump Suction 23MOV-57 Wil not be tested as lines are water scaled by suppression chamber wamt.
Forus) 23MOV 58 227A Cosa Spray - Pump 14MOV-7A Wdt not be tested as line is water scalmi by suppression chamber water.
Suction Forus) 2278 Core Spray - Pump 14MOV-78 WM not be tested as line is water sealed by suppression chamber water.
Swtion (Torus) 228 Condensate to Torus 33CND 102 WM not be tested as line is water sealed by suppression clanter water.
D et*
6' w\
@ ,,in.,143 Pl;tli
~ . . _ _ _ _ _ .-
l 6.19 POST CCIDENT SAMPLING PROC M A program shall be established. implemented, and maintained which will ensure the capability to obtain and analyse reactor coolant , radioactive iodines and particulates in plant gaseous effluents , and containment atmosphsre samples under accident conditions.
the following: The program shall include A) Training of personnel.
3)
C) procedures for sampling and analysis, provisions for maintenance of sampling and analysis N'#
roseer D l
I l
Amendment No. 1 '
258e
Jwa s em a m_ma -
- e.
INSERT D:
"6.20 Primarv Containment Leakaae Rate Testino Proaram A program shall be established to implement the leakage rate testing of the Primary Containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, 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 free 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 packing) are not subjected to test pressure, the portions of the valve not exposed to test pressure shall bc ::9bjected to leakage rate measurement during regularly scheduled Type A te', ting. A list of these valves, the leakage rate measurement method, and the accepance criteria, sha:t 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.
C. The leakage rate acceptance criteria are:
- 1. Primary containment leakage rate acceptance criteria is i 1.0 L,. During ,
unit startup following testing in accordance with this program, the leakage rate acceptance criteria are 10.60 L, for the Type B and Type C tests and 10.75 L, for the Type A tests; l 2. Airlock testing acceptance criteria are:
- a. Overall airlock leakage rate is 10.05 L, when tested at A P,,
- b. For each door seal, leakage rate is 5120 scfd when pressurized to j A P,.
- 3. MSIV leakage rate acceptance criteria is 111.5 scfh for each MSIV when tested at 125 psig D. The provisions of Specification 4.0.8 do not apply to the test frequencies specified
- in the Primary Containment Leakage Rate Testing Program.
! E. The provisions of Specification 4.0.C are applicable to the Primary Containment i Leakage Rate Testing Program."
-. -..--.-.. . .-- . . . - . ~ . - . . . . - . . - - . - . . - - . . . . - - . - - - - . . - - - - -. - - - -
JAFNPP -
Go , L e. l . \ G3, " t% fem.m,_.
Rey \ ^
i hd Ga4- M 6'- h T '
7.0 REFERENCES
P~ y- l' J Je3 @ 6d v ITIS I ;
E. Janssen, " Multi-Rod Bumout at Low Pressure," ASME Paper (9) C.H. Robbins, " Tests of a Full Scale 1/48 Segment of the (1) Humbolt Bay Pressure Suppreselon Conseinment," GEAP-3596, 62-HT-26, August 1962.
November 17,1960.
K.M. Becker, "Bumout Commons lor Flow of BolEng Water in (2) (10) "Nudeer Safety Program Annual Progress Report for Pedod Ver$ cal Rod Clusters," AE-74 (Socidickn, Sweden), May 1962.
EndnD D ecember 31,1986, Progress Report for Podod Ending l December 31,1986 ORDE.4071."
(3) FSAR Socdon 11.2.2.
FSAR SecGon 4.4.3. (11) Secton 5.2 of the FSAR.
(4) 1.M. Jacobs,"Reuebluty of Engineered Salsty Features as a (12) TID 20583, "Leekage Characteristics of Steel Containment
-(5) Vessel and the Analyels of Leake0s Rate Determinadons " j Function of Testing Frequency," Nuclear Safety, Vol. 9. No. 4, (13) Tactinical Safety Guide, " Reactor Containment Leeka0e Testing and Survemance Requirements," USAEC, Division of Safety (6) Deleted Standants, Revised Draft December 15,1966.
1.M. Jacobs and P.W. Madott, APED Guidelines for Determinin0 (7)
Safe Test intesvals and Repair Times for EnWneered Safeguents (14) Secdon 14.6 of the FSAR.
- Apre 1969.
(15) ASME Boller and Pressure Vessel Code, Nucteer Vessels, Section Ill. Maximum atomobie intemel pressure is 62 psig.
(8) Bodege Bay Preliminary Hazards Report, Appendx 1, Docket 50 205, December 26,1962.
(16) '$U CFR 50.54, Appendx J, " Reactor Containment Testing tRequimments "
(17)60 CFR 50, Appendix J. February 13,1973. Da.k k &
L /O q r=-R Pn-l Sb Appw c/ n i
fr* ~ ) /d* N Gak.a n + Le L ye. T-s h ap & uh. k, Go le0 Po s<c R ackes o, M B - Per % re L.D Reyv,< w Q,'/~/L M dak- Oc 4 Ler .x ,19 95_
Amendment No.1)lb )<- 285
f l
l Attachment IV to JPN-96-016 1
l APPENDIX J OPTION B IMPLEMENTATION PLAN l
1 i
(JPTS-96-003) i 1
i i
8
)
l New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR 59
~ .._ __ _ . . - - _
Attachment IV to JPN-96-016 APPENDIX J OPTION B IMPLEMENTATION PLAN Page 1 of 5 l!iTRODUCTIO_N Option B of 10 CFR 50, Appendix J (Option B) provides a performance based approach for 8eakage rate testing of primary containm,nt. This action improves the focus of the regulation by eliminating prescriptive requirements that have been determined to be marginal in safety.
Opti on B allows for test intervals to be established based on system and component performance and provides for greater flexibility for cost effective implementation methods of regulatory safety objectives.
This plan outlines how the Authority willincorporate Ootion B into the Primary Containment Leakage Rate Testing Program (the Program) for Type A, B and C testing at the FitzPatrick plant. The Authority will comply with the requirements contained within References 1 through 4.
COMPONENT LEAKAGE LIMITS Fitzpatrick will set administrative limits for each Appendix J component and develop the procedures for changing them. The existing plant administrative limits will be reviewed and compared against consistent limits set by the FitzPatrick Maintenance Rule Expert Panel.
A component's measured leakage is compared against its administrative limit to determine whether the As-Found LLRT passed or failed on a performance basis. The expert panel will review and approve administrative leakage rate limits since the proper setting of these limits is extremely important under the performance-based rule. Comparison of a components As-Found leakage against the administrative limits will determine if a test passed or failed , thus, the values chosen will affect e'ach component's Type B or C testing frequency.
Two limits, a warning limit and an alarm limit, will be specified for each component. A component should be repaired if the As-Found leakage rate is above the warning limit, but below the alarm limit. If repaired, an As-Left test will be conducted. The As-Found test is not counted as a performance failure. If a component's leakage rate is above the alarm limit, then the component shall be repaired. The component will be retested after the repair. The As-found test is counted as a performance failure. This scheme allows for a low leakage setpoint to trigger component repairs so as te maintain containment in good condition. It also allows for the alarm limits to be set high enough that a Type B or C As-Found test need not be counted as a failure unless the component is found in a seriously degraded condition.
Although administrative limits are used to maintain the containment in good condition, it should be noted that the sum of the As-Left Maximum Pathway Leakage Rates for all Appendix J barriers must be less than 0.6 La per plant Technical Specifications (TS) prior to entering a mode requiring primary containment integrity. In past instances where leakage from one or more components have exceeded administrative limits, and correcting this condition would have either been very difficult or costly, a total containment leakage evaluation was performed and documented. If the evaluation concluded that the additional leakage posed no significant safety impact, and the TS limit of 0.6La was not exceeded, the component (s) was(were) allowed to continue to leak in excess of the individual valve leakage
I l
Attachment IV to JPN-96-016 APPENDlX J OPTION B IMPLEMENTATION PLAN Page 2 of 5 administrative limit until repairs could be made. The test is still considered to be a failure because the administrative alarm limit was exceeded. The Authority reserves the option to continue use of this criteria when the alarm limit is exceeded, only on a critical as needed basis.
BUILDING PERFORMANCE BASELINES / ESTABLISHING TEST FREQUENCIES Tvoe A Test Type A testing procedures will be revised per the new Option B requirements, and shall be performed during a period of reactor shutdown at a frequency of at least once per 10 years based on acceptable performance history. Acceptable performance history is defined as completion of two consecutive periodic Type A tests where calculated performance leakage rate was less than 1.0 La. Elapsed time between the first and last tests in a series of consecutive satisfactory tests used to determine perforrnance shall be at least 24 moriths.
Option B allows for reviewing performance history with several options to determine if past Type A tests were satisfactory.
- a. As-Found Type A test results can be compared to 1.0 La rather than the previous 0.75 La criteria.
- b. Leakage savings (repairs / adjustments) from type B and C testable pathways which were added as penalties to the As-Found Type A test can be subtracted when reviewing previous Type A test results.
- c. The Type A test UCL from previous Type A tests may be recalculated using the Mass 1 Point Methodology described in ANSI /ANS 56.8-1994.
The Authority has reviewed Type A test results for the FitzPatrick plant as compared to the new requirements / criteria to establish a test frequency for the Primary Containment I Integrated Leak Rate Test (ILRT). It has been determined that the two most recent As-Foaad Type A tests (1990 and 1995) are below the 1.0 La criteria. Leakage savings (repairs /ar'justments) from Type B and C testable pathways which were added as additions to the As-Fowd Type A test were not subtracted. Based on this, Fitzpatrick will implement the 10 year Type A test frequency based on the criteria set forth in the new rule. The Type A test interval may be extended by up to 15 months, however, this option will only be used in cases where refueling schedules have been changed to accommodate other factors.
1 l
Attachment IV to JPN-96-016 APPENDIX J OPTION B IMPLEMENTATION PLAN Page 3 of 5 l
Appropriate administrative controls have been developed such that prior to initiating a Type A I test, a visual examination shall be conducted of accessible interior and exterior surfaces of )
the containment system for structural problems which may afkct either the containment '
structure integrity or the performance of the Type A test. Dese containment inspections will also be conducted during two other refueling outages before the next Type A test.
Therefore, FitzPatrick will perform these examinations at least three times every ten years, regardless of the Type A test schedule.
Tvoe B and C Test i
l The Authority will develop a procedure for building and documenting Type B and C testing '
performance baselines for the FitzPatrick plant. This procedure will be used to ensure that a consistent criteria is applied to establish component baseline performance and their subsequent testing frequencies. The Authority will develop bases for test frequencies based upon performance of leakage tests that meet the requirements of Option B and approved exemptions, in addition historical performance, considerations such as service life, environment, design, system application, special service conditions, and safety impact / risk l from failure will be reviewed / evaluated in determining test frequency. The component's performance history will determine its test interval. l l
The Authority will compile the required leak rate historical data and continue to update this I data with the most current As-Found leak rate data. The performance history of each component will be evaluated against the alarm limit to rate component performance over the last two refuel outages.
Type B components which are determined to have a performance rating of unknown, poor, or improving, will require a 30 month test frequency. A rating of good or excellent allows for up to a 120 month test interval. The component will be evaluated to determine if it is a member of a group of components subject to the same common mode failure mechanisms, if so, then the test intervals of all members of that group will be staggered, such that some percentage of those components are tested periodically. The date of the next test may be earlier than required by the baseline interval for this reason. The Authority intends to place good or excellent performing Type B components on a 120 month test interval. The test frequencies of similar/ grouped components will be staggered to ensure that a percentage of components are tested periodically.
Type C components which are determined to have a performance rating of unknown, poor, or improving, will require a 30 month test frequency. A rating of good or excellent allows for up to a 60 month interval.
Per NRC Regulatory Guide 1.163, the NRC does not endorse extended test intervals of greater than 60 months for Type C tests. Further, the Regulatory Guide states that Type C tests for Main Steam and Feedwater isolation valves and containment purge and vent valves, should be limited to 30 months with consideration given to operating experience and safety significance. The Authority intends to fully comply with this guidance by performing Type C tests on Main Steam, Feedwater, and Containment Vent and Purge isolation valves at a 30 l
l l
Attachment IV to JPN-96-016 APPENDIX J OPTION B IMPLEMENTATION PLAN Page 4 of 5 month interval with consideration given to operating experience and safety significance.
Type B and C testing intervals may be extended by 25 percent of the interval, not to exceed 15 months, however, this option will only be used in cases where refueling schedules have been changed to accommodate other factors.
The Authority will place the primary containment airlock on a test frequency of at least once per 30 months. Airlock door seals will be tested within seven days after each containment access when primary containment integrity is required. For periods of multiple containment entries where the airlock doors are routinely used for access more frequently than once every seven days, door seals may be tested once per 30 days during this time period.
EelLURES, REPAIRS / ADJUSTMENTS, CORRECTIVE ACTIONS Failures, repairs / adjustments, and corrective actions for Type A, B, and C testing results will be evaluated through the Deviation Event Report process.
If Type A performance leak rate test results are net acceptable, then a determination will be performed to identify the cause of unacceptable performance and determine appropriate corrective actions. Once the cause has been determined, and corrective actions have been completed, acceptable performance should be reestablished by performing a Type A test within 48 months following the unsuccessful Type A test. Following a successful Type A test, the surveillance frequency may be returned to once per 10 years.
Type B or C component failures discovered during performance of the Type A test will be considered as failure of a Type B or C test for purposes of cause determination and corrective action. This includes failures of type B and/or C components that were not previously identified by a Type B or ^ test.
Type B and C component failures will require that testing frequency be set at the baseline test interval of 30 months. A cause determination will be performed and corrective actions identified that focus on those activities that can eliminate the identified cause of failure and prevent recurrence. Once the cause determination and corrective actions have been completed, acceptable performance should be reestablished and the testing frequency returned to the extended interval in accordance with the NEl 94-01 guidance.
l In addition to the periodic As-Found Type B and C test, an As-Found test shall be performed i prior to maintenance, repair, modification, or adjustment activity if the activity could adversely affect the penetration leak tightness. An As-Left Type B or C test shall be performed following those activities, unless engineering analysis shows reasonable assurance that such work does not affect the leak tightness of the penetration and that it can still perform its intended function. Specifically for Type C tests, an alternative method or analysis can be used to provide reasonable assurance that such work does not affect a valve's leak tightness and a valve will still perform its intended function, if As-Found and As-Left Type B and/or C results are both less than the allowable administrative limit, a change in testing frequency is not required. If the results are unacceptable, testing shall continue at initial test intervals until adequate performance history is reestablished.
1
,nax .._
Attachment IV to JPN-96-016 APPEriDIX J OPTION B IMPLEMENTATION PLAN Page 5 of 5 TECHNICAL CRITERIA AND TESTING METHODOLOGY INTERPRETATION l l Changes to the leak rate testing program will be required regarding testing methodology and procedural requirements under Option B. The technical details / criteria and testing methodology as described in ANSl/ANS 56.8-1994 will be used in updating the FitzPatrick Type A, B, and C leak rate testing program.
( REFERENCES l
l 1. Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995
- 2. Option B of 10 CFR 50 Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors (60 FR 49495)
- 3. NEl 94-01, Industry Guideline for implementing Performance-Based Option of 10 CFR l 50 Appendix J, Rev. O, dated July 26,1995 i l i
- 4. ANSI /ANS-56.8-1994, Containment System Leakage Testing Requirements i
I l
l i
l l
l
Attachment V to JPN-96-016 LIST OF COMMITMENTS
- Commitment No. Description Due Date i
- JPN-96-016-01 Perform the appropriate actions to July 1,1996 or prior qualify the Standby Gas Treatment to implementation of a fan motors and the GE Motor Control 1,5 percent per day Centers for a 1.5 percent per day allowable leakage rate.
primary containment leakage rate.
JPN-96-016-02 Revise plant procedures to incorporate July 1,1996 or prior a soap bubble test during ILRT for the to implementation of a
< 17 containment isolation valves listed 1.5 percent per day in Table 1, utilizing the acceptance allowable leakage rate.
criteria of zero bubbles.
4 1
l 1