ML20031B946
| ML20031B946 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 09/28/1981 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20031B939 | List: |
| References | |
| RTR-NUREG-0313, RTR-NUREG-313, TASK-A-42, TASK-OR GL-81-03, GL-81-3, NUDOCS 8110060258 | |
| Download: ML20031B946 (11) | |
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1 ATTACHMENT I PROFOSED TECHNICAL SPECIFICATION CHANGE RELATED TO MATERIAL SELECTION AND PROCESSING FOR BWR COOLANT PRESSURE BOilNDARY PIPING 1
POWER AUTHORITY OF THE STATE OF NEW YORK i
JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 i
(
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P O$000333-PDR.
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JAFNPP 3.6 (cont'd) 4.6 (con't) 4.
Except as specified in 3.6.C.3 above, the reactor coolant water shall not exceed the following limits with steam-ing rates greater than or equal to 100,000 lb/hr and during reactor shut-downs.
Conductivi ty 5 pmho/cm Chloride ion 0.5 ppm 5.
If Specification 3.6.C cannot be met, the reactor shall be placed in a cold condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D.
Coolant Leakage 4.6.D Coolant Leakage 1.
Anytime irradiated fuel is in the Reactor coolant leakage rate inside the reactor vessel and the reactor coolant primary containment shall be established temperature is above 2120F, the reactor once/ day utilizing the Equipment and Floor 4
coolant leakage into the primary con-Drain Sump Monitoring Systems.
tainment shall be limited to:
1 a.
C gpm unidentified leakage j
b.
2 gpm ino: ease in unidentified leak vge within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
(This limitation shall apply only after a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at oper-ating pressure.)
i c.
The total reactor ceolant leakage into the primary containment shall not exceed.25 gpm.
4 2.
With ar,y reactor coolant system leakage greater than any one of the limits speci-fied in a. or c. above, the leakage rate shall be reduced to within the limits 141
4 JAFNPP
~
i 3.6 (cont'd) 4.5 (cont'd) i within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or the reactor shall be in at least the hot standby condition within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold condition within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.
If the increase in unident3 fied leakage 4
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as speci fied in 3.6.D.I b is exceeded, the source of the leakage shall be identified within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or the reactor shall be in at least hot standby condition within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold condition a
within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.
The-following reactor coolant system leakage detection systems shall be operable during reactor power operation:
a.
Drywell Sump Monitoring System (equipment drain sump monitoring and floor drain sump monitorino),
b.
Drywell Continuous Atmosphere (particulate) Radioactivity Monitoring System, and c.
Drywell Continuous Atmosphere (gaseous) Radioactivity Monitoring System.
141a
3.6 (cont'd)
JAFNPP 4.6 (cont'd) l 5,
with only two of the leakage j
detection systems operable (3.6.D.4), operation may l
continue for up to 30 days provided grab samples of the drywell atmosphere are obtained 3.
Drywell Continuous Atmosphere and analyzed at least once per Radioactivity Monitoring System j
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the required gaseous instrumentation shall be funct-1 or particulate monitoring system ionally tested and calibrated as is inoperable; otherwise, be specified in Table 4.6-2.
in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in cold shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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Amendsunnt No.
142
s JAFNPP 3.6 (cont'd) 4.6 (cont'd) 3.
An augmented inservice inspection program is also required for Reactor Coolant System Pressure Boundary piping and weld metal of 304 or 316 austentic stainless steel with greater than 0.035 per cent carbon. The augmented program shall be performed on ASME Code class 1 piping requiring inspection as per 1974 Edition, Summer 1975 Addenda of Section i
XI of the ASME Boiler and Pressure Vessel Code.
The augmented inspection program shall be as described in NUREG-0313, Rev. 1, dependent on j
pipe classification of service or non-service i
sensitive as follows-l l
a.
The stainless steel 28-inch and 22-inch recirculation system piping, and the stain-I l.
less steel piping in the Residual lleat j
(
Removal and Reactor Water Cleanup piping
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are classified as non-service sensitive.
I I
b.
The bypass of the discharge valve in the l
main recirculation loops, and the stainless I
);
steel portion of the core spray system, and the control rod drive hydraulic return piping excluding safe-ends are classified as service sensitive.
c.
The recirculation riser piping excluding safe-ends is classified as non-service sensitives however, itshall be inspected on an initial 40-month schedule. If after three successive inspections, no indications of IGSCC are found, the programshall be performed on an 80-month schedule.
G.
Jet Pumps G.
Jet Pungs Whenever the reactor is in the whenever there is recirculation flow with the utor startup/ hot standby or run modes, fn the startup/ hot standby or run modes, jet pua,,
all jet pumps shall be operable, operability sl:all be checked daily be verifying ti-If it is determined that a jet pump the following conditions do not occur simultaneour's is inoperable, the reactor shall be placed in a cold condition within 24 l
144a hours.
m
3.6 and 4.6 BASES (cont'd)
JAFNPp not required to be operable (reactor after the system is in service.
coolant temperature 212 F and the This inspection should reveal reactor vessel vented or the reactor problem areas should they occur vessel head removed). permitting before a leak develops.
physics testing and operator training under these conditions Several locations on the main steam would not place the plant in an lines and feedwater lines are not unsafe condition.
restrained to prevent pipe whip in the event of pipe failure at these F.
Structural Integrity locations. The physical layout within the drywell precludes i
p A pre-service inspection of the com-restraints at these points.
ponents listed in Table 4.6-1 will Unrestrained high stress areas have 1
be conducted after site erection to been identified in these lines where l
assure the system is free of gross breaks result in pipe whip l
defects and as a reference base for such that the pipe could impact the L
later inspections. prior to primary containment wall. Augmented operation, the Reactor Coolant inservice inspection of these weld System will be free of gross location shall be performed during j
defects.
In addition, the facility each inspection period.
has been designed such that grose defects should not occur throughout An augmented inservice inspection shall also be i
life. The inspection program given performed on piping within the Reactor l
in Appendix F of the FSAR is based Coolant Boundary which is of type 304. The on the ASME Code,Section XI. for recirculation system,(except the risers),and in-serice inspection which was parts of the core spray, residual heat removal, followed except where accessibility and reactor water clean up systemshall be examined for inspection was not provided.
as specified by NUREG-0313, Rev. 1, July 1980.
The applicants recognize the The recirculation risers shall be inspected as importance of inspection of those described in NUREG-0313, Rev. 1, except that an areas which are presently not initial forty month schedule ahall be employed.
accessible and will study and The stainless steel safe-ends are manufactured implement, if practicable, new means from low carbon steel and are not susceptible to include those areas within the tointergranular stress corrosion cracking.
inspection program. This inspection provides further assurance that gross defects are occurring Amendment No.
151
3.6 and 4.6 BASES (cont'd)
JAFNPp In addition, extensive visual in-spection for leaks will be made periodically on critical systems.
The inspection program specified encompasses the major areas of the vessel and piping systems within the drywell. The inspection period is based on the observed rate of growth of defects from fatigue studies sponsored by the AEC.
l These studies show that it required thousands of stress cycles at stresses beyond any expected to occur in a Reactor Coolant System to propagate a crack. The test Amendment No.
153a
a f
7.O REFERENCE,S JAFNPP (1)
E. Janssen,
" Multi-Rod Burnout at (9)
C.H. Robbins, " Tests of a Full Low Pressure," ASME Paper 62-HT-26, Scale 1/48 Segment of the Humbolt August 1962.
Bay Pressure Suppression Containment.' GEAP-3596, November (2)
K.M. Backer, " Burnout Condicions 17, 1960.
for Flow of Boiling Water in Verti-cal Rod Clusters," AE-74 (10) " Nuclear Safety Program Annual (Stockholm, Sweden), May 1962.
Progress Report for Period Ending December 31, 1966, Progress Report (3) FSAR Section 11.2.2.
fer Period Ending December 31, 1966, ORNL-4071."
(4) FSAR Section 4.4.3.
(11) Scction 5.2 of the FSAR.
(5)
I.M. Jacobs, " Reliability of Engi-neered Safety Features as a Func-(12) TID 20L93, "Leakaoe Characteristics tion of Testing Frequency," Nuclear of Steel Containment Vessel and the Safety, Vol. 9, No. 4, July-A gust Analysis of Leakage Rate Determi-1968, pp 310-312.
naticas."
(6)
Benjamin Epstein, Albert Shiff, (13) Technical Safety Guide, " Reactor UCRL-50451. Improving Availability Containment Leakage Testing and and Readiness of Field Equipment Surveillance Requirements," USAEC, Through Periodic Inspection, July Division of Safety Standards, 16, 1968, p. 10, Equation (24),
Revised Draft, December 15, 1966.
Lawrence Radiation Laboratory.
f(4) Section 14.6 of the FSAF.
(7)
I.M. Jacobs and P.W. Mariott, APED Guidelines for Determining Safe (16) ASME Boiler and Pressure Vessel Test Intervals and Repair Times for Code, Nuclear Vessels,Section III.
Engineered Safeguards - April 1969.
Maximum allowable internal pressure is 62 psig.
(8)
Bodega Bay Preliminary Hazards Re-port, Appendix 1, Docket 50-205, (16) 10CFR50.54, Appendix J, Reactor Con-December 28, 1962.
tainment Testing Requirements."
(17) 10CFR50, Appendix J, February 13, 1973.
(18) NUREG-0313, Rev. 1, " Technical Report on Material Selection and Processing Guidelines for BWR Coolent Pressure Boundary Piping",
285 July 1980.
b ATTACHMENT II SAFETY EVALUATION RELATED TO u
2 MATERIAL SELECTION AND PROCESSING FOR BWR COOLANT PRESSURE BOUNDARY PIPING i
POWER AUTHORITY OF THE STATE OF NEW-YORK JAMES A.
FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 i
i i
i 1
Srction I - Description of Modification The proposed changes to the James A. Fitz?atrick (JAP)
Technical Specifications are included as Attachment I, and amend Sections 4.6.D (pages 141, 141a, 144a), 3.6 and 4.6. BASES (pages 153, 153a, 285).
Section II - Purpose of Modification The purpose of modifications is ta incorporate appropriate augmented inservice inspection requirements specified in the Commission's February 26, 1981 letter (Generic Letter 81-04) regarding the implementation of NUREG-0313, Revisien 1 entitled
" Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping (Generic Task A-4 2). "
The proposed amendment further incorporaten changes to bring the JAF Technical Specifications into conformance with the model technical specifications included with this letter.
Tne Authority has previously submitted to the Commission tne results of our review of Reactor Coolant System pressure boundary materials to determine the extent of compliance with the guidelines of NUREGe0313, Revision 1 (Reference (c).
These proposed enange implement tne findings of tnat review.
Secton III - Impact of the Change Tnese changes to the JAF Technical Specifications will not alter the conslusions of either the FSAR or SER accident analysis.
Section IV - Implementation of the Modification The changes as proposed will not impact the Fire Protection Program at JAF.
Additional inservice inspections required to implement Section IV of NURBG-0313, Revision 1 will impact the ALARA program tnrough increased occupational radiation exposure.
Section V - Conclusion Tne incorporation of these modifications:
a) will not change the probability nor the consequence of an accident or malfunction of equipment important to safety as previously evaluated in tne Safety Analysis Report; b) will not increase the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report; l
and c) will not reduce tne margin of safety as defined in the basis for any i
Technical Specificaton, and d) does not constitute an unreviewed j
safety question.
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S1ction VI R2ferences 1
(a)
JAF FSAR (b)
JAF SER (c)
PASNY July 31, 1981 letter, J.P. Bayne to T.A. Ippolito, same subject (d)
NUREG-0313, Revision 1, " Technical Report on Materi l a
Selection and Processing Guidelines for BWE Coolant Pressure Boundary Piping", July, 1980 4
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