ML20093H666
| ML20093H666 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 10/09/1984 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20093H559 | List: |
| References | |
| NUDOCS 8410160344 | |
| Download: ML20093H666 (4) | |
Text
- 17 p
.g V
ATTACHMENT fI TO JPN 64 PROPOSED TECHNICAL SPECIFICATION CHANGE RELATED TO REACTOR COOLANT LEAKAGE DETECTION NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 l
l 8410160344 841009 gDRADOCK 05000333 PDR
1
.J fWWl'I' 3.6 (cont'd) 4.6 (cont'd)
~
4.
Except as specified in 3.6.C.3 above, the reactor coolant water shall not exceed the following limits with steaming rates greater than or equal to 100,000 lb/hr and during reactor shutdowns.
Conductivity 5 4mho/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 vessel and the reactor coolant Reactor coolant leakage rate inside the primary containment shall1be temperature is above 212 F, the reactor coolant leakage into the primary con-monitored and recorded once every tainment shall be limited to:
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> utilizing the Primary Con-4 tainment Sump Monitoring System (equipment drain sump monitoring and 5 gpm unidentified Icakage floor drain sump monitoring).
a.
b.
2 gpm increase in unidentified leakage 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.)
The total reactor coolant leakage c.
into the primary containment shall not exceed 25 gpm.
2.
With any reactor coolant system Icakago greater than any one of the limits speci-fied in 3.6.D.1.a or 3.6.D.l.c kove, the leakage rate shall be reduced
.i. thin these limits Amondmont No. JT, JAf 141
..:z.
--7--
- a~
~ ^ ^:
$6 k
a cx
, q:: _ -.
]
~-
13.6 - (cont'd)
~
4.5-[ cont'd) 1:
- b 3
y 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.
i
~
(
be_in at least the hot standby con-dition within the following.12-hours 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' unidentified leak-age as specified in 3.6.D.l.b,is ex-ceeded,-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 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 primary Containment Sump Monitoring System (Equipment Drain Sump Monitoring and Floor Drain Sump Monitoring) and the Primary Containment Atmosphere Mon-itoring System (Gaseous and Particulate) chall be operable during reactor power operation.
~
4 1
I Amendment No.
141a
~
+
l
'.l
JAFNPP D
3.6 (cont'd) 4.6 (cont'd)
~
5..
With the Primary Containment Sump 3.
Drywell continuous Atmosphere Monitoring System (Equipment Radioactivity Monitoring System-Drain Sump Monitoring or instrumentation shall be' functionally Floor Drain Sump Monitoring) tested and calibrated as specified inoperable,.r~estore the.
in Table 4.6.2.
g system to operable status
'within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or immediately initiate a and bo-in.q orderly shutdown 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 within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
6.
With the Primary Containment Atmosphere Radioactivity Mon-itoring System (gaseous) or the Primary Containment Atmosphere Radioactivity Monitoring System (particulate) inoperable, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Otherwise be 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 following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1 4
5 Amendment No.j)(I ks 142 j
2
7
'i-I
' ATTACHMENT II TO^JPN-84-64 SAFETY:EVALUTION, RELATED TO
-REACTOR COOLANT LEAKAGE DETECTION s'
NEW YORK POWER AUTHORITY JAMES A.
FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 i
e l
I
c a
-9 7
/-
- i Section I - Description of Changes The proposed changes tar the James A. FitzPatrick (JAF)
' Technical Specifications-are included as_ Attachment I, and amend Section 3.6.D-(Coolant ~ Leakage) on pages 141,-141a and 142.
The
~
changes impose stricter limitations on leakage rates from unidentified: sources inside the primary containment.
Section'II - Purpose of Changes
'During'a'recent FitzPatrick plant outage (started March 1, 1984), the Authority treated eleven recirculation system welds with Induction Heating Stress Improvements.(IHSI) methods to demonstrate
^
the-benefits of IHSI as an IGSCC-(Inter Granular Stress Corrossion Cracking) countermeasure.
During pre-IHSI inspections, a single weld showed linear crack indications (Reference g).
In Reference (f), the NRC staff informed the Authority that stricter limitations were being imposed on leakage rates from unidentified sources for boiling water reactors-where inspections have uncovered evidence of IGSCC.
Reference (f) further requested that the Authority submit an updated version of-proposed Technical i
specifications previously proposed to address Revision 1 to NUREG-0313 (Reference.e).
The changes included as part of this amendment application respond to this request.
Section III - Impact of the Change The changes to the FitzPatrick Technical Sepcifications will
' not alter the conclusions of either the FSAR or SER accident analysis.
The Authority considers that this proposed amendment can be classified as not likely to involve significant hazards considerations since the change constitutes additional limitations l
not' presently included in the Technical Specifications.
In l
particular, these new-limitations will inpose additional limiting conditions for operation.
This is clearly in keeping with example (ii) included in Federal Register, Vol. 48 No. 67 dated April 6, 1983 page 14870,- (Examples of Amendments that are Considered not
. Likely.to Involve Significant Hazards ~ Consideration'which states:
"A change that constitutes an. additional limitation, restriction, or control not presently included in the technical specifications:
for example, a more' stringent surveillance requirement.")
j 1
v--3 w
-,,vy,<,
y, m q--
e i, m v -y
-c,-w y-
,--.,--ete,,,,-+-----
-,--y,
,-m,%
,-,,--w+ett
--me-
- -w-r-+-- * + +
+6-*wr*m-*---=ar-**
m~i
'r------e-~-
-- e
a Section IV - Implementation of the Change Implementation of these changes, as proposed, will not impact
.the ALARA or fire protection programs at FitzPatrick, nor will the changes impact the environment.
Section V - Conclusion The 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 the Safety Analysis Report; b) will not increase the possibility for an accident or malfunction of a different type than any evaluated l
previously in the Safety Analysis Report; c) will not_ reduce the margin of safety as defined in the basis for any Technical Specification; d) does not constitute an unreviewed safety question, and e) involves no significant hazards consideration, as defined in 10 CFR 50.92.
Section VI - References (a)
James A.
FitzPatrick Nuclear Power Plant Final Safety Analysis Report (b)
Safety Evaluation by the Division of Reactor Licensing, U.S. Atomic Energy Commission in the Matter of Power Authority of the State of New York, James A.
FitzPatrick Nuclear Power Plant dated March 4, 1970 ao amended.
(c)
PASNY July 31, 1981 letter (JPN-81-54), J.P.
Bayne to T.A.
Ippolito, regarding implementation of NUREG-0313, Revision 1.
(d)
NUREG-0313, Revision 1, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping", July 1980.
(e)
PASNY September 28, 1981 letter (JPN-81-76), J.P. Bayne to l
T.A.
Ippolito, regarding proposed changes to the Technical Specifications related to the implementation of NUREG-0313, Rev.l.
t l
(f)
NRC April 13, 1984 letter, D.B. Vassalo to J.P. Bayne, regarding intergranular stress corrosion cracking (IGSCC) in the recirculation piping system.
(g)
NYPA March 9, 1984 letter (JPN-84-16), J.P. Bayne to D.B.
f Vassallo, regarding recirculation piping flaw indication.
t
r g -...
-(h) -NYPA letter (JAFP-83-0769), C.A. McNeill to T.E. Murley, rdated July 22, 1983.
(i)- 'PASNY~ January 6, 1978 letter (JNRC-78-1),-G.T. Berry to R.W. Reid,- regarding review of reactor..ccolant system pressure. boundary piping susceptible to stress corrosion
- cracking.
(j)
NRC February 26, 1981 letter, D.G.;Eisenhut.to all BWR licensees regarding implementation of NUREG-0313, Revision 1-(Generic Letter No. 81-04).
(k)' NRC September 29, 1977 letter regarding use of, type or.304 and 306 austenitic stainless steel'in reactor coolant pressure boundary.
i.
a 1
.~.......-.