ML20042F623

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Proposed Tech Specs Reflecting Mods to Traversing Incore Probe Sys Scheduled for Installation During 1990 Refueling Outage
ML20042F623
Person / Time
Site: FitzPatrick Constellation icon.png
Issue date: 05/04/1990
From:
POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20042F622 List:
References
NUDOCS 9005090160
Download: ML20042F623 (9)
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CORRECTED TECHNICAL SPECIFICATION PAGES New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR 59 ggg50ggobOb [3 P

ar JAFNPP TABLE 3.7 (Sh.6 of 15) _

PRIMARY CONTAINMENT ISOLATION VALVES -

' CONTAINMENT PENETRATION VALVE ISOLATION CLOSE TIME NORRhAL -REMARKS PENETRATION FUNCTION NUMBER- S!GNAL (SEC) (5) - STATUS (7) 31Bd Drywell 27SOV-135D .A,F,R,Z N/A Open From elev. 293' to atmospiere Radiation Monitors '

sample 27SOV-135B . A.F,R,Z N/A Open Note 12.

(Suction) 35B Traversing 07SOV-104A. A.F.R N/A Open Notes 8,14, in e Probe "A* 07EV-104A R N/A Open Notes 14,15.

35C Traversmg 07SOV-104C A.F.R N/A Open Notes 8,14 In-core Probe "C* 07EV-104C R N/A Open Notes 14,15.

35D Traversing 07SOV-1048 A.F.R N/A Open Notes 8,14.

In-core -

Probe *B" 07EV-104B R N/A Open ' Notes 14,15.

Amendment No. [, f.1[,1[

203

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JAFNPP

,-. Notes For 'lable 3.71

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11. Valve 20AOV.95 opens during pump out of the drywell equipment sump. Automatic 'f Isolation signals A and F override an open signal that might be present for sump pump out. .
12. Radiation monitors used for sampling lodine, particulate, and gaseous are as follows:

Radiation .

Monitors Sample i

17 RM 101 A lodine

.17.RM-1018 17.RM 102A Particulate 17.RM.102B 17 RM.103A Gaseous 17 RM.103B

13. Isolation signais A, F, and Z may be manually overridden by keylock switch on the i Monitoring and Analysis Panel (MAP) located in the relay room.
14. Traversing in-core Probe (TIP) penetrations are isolated by a guide tube and valve assembly which includes a solenoid operated ball valve and an explosive shear valve <

designed to sever and seal the TIP tubing and TIP detector helix.

15. The explosive shear valves are not normally actuated and require replacement parts and maintenance activity in order to open the valves following actuation.

Amendment No. 46, p, ipd 209

JAFNPP' s TABLE 4.7-2 EXCEPTION TO TYPE C TESTS -

The following penetrations are excepted from Type C testing regthvir.wiE VALVE LOCAL LEAK RATE TEST PERFORMED CONTAINMENT - PENETRATION FUNCTION NUMBER-PENETRATION Main Steam - . 29AOV-80A The inboard valves wil be tested in the reverse direction. Pressure wEl be 7A apphed between the isolation valves and leakage enessured. A water seal 78 - 29AOV-808 ~

29AOV-80C of 25 psig wil be used on the inboard valve to determine the outboard 7C valve's leak rate. (limit 11.5 SCFH st 25 psig.) ~

7D ' 29AOV-80D

' 29AOV-86A 29AOV-86B 29AOV-86C 29AOV-860 27AOV-112 These valves wil be tested in tise reverse a;.Mi.s 25 Drywell Purge Inlet (Airand/or 27AOV-131A Nitrogen) 27AOV.131B

- 27AOV-113 These valves wil be tested in the reverse direction.

26A'& 26B Drywell Purge Inlet (Air and/or 27MOV-122 Nitrogen)

Various Wil not be tested as lines are sealed by process fluid.

30A lostrumentation 07EV-104A This valve is an explosive shear valve vnkh cannct be Type C tested.

35B Traversing Incro Probe "A*

Amendment No.f f. ty6,1[.1/5 211

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i REVISION 1 TO SAFETY EVALUATION 6

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1 New York Power Authority .

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No 50333 ,

DPR 59 .

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i Attachment 11

- SAFETYEVALUATION REVISION 1 Page 1 of 3 1, DESCRIPTION OF THE PROPOSED CHANGES f The proposed changes to the James A. RtzPatrick Technical Specifications revise Table 3.7-1,

" Primary Containment Isolation Valves,' and Table 4.7 2,

  • Exception to Type C Testc." The l ,

changes are as follows:

Table 3.71, paDe 203 ,

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1. Delete penetrations X 35A and X 35E. i
2. Revise valve numbers as follows:

Penetration From valve number To valve number t 358 07NM 104B 07SOV 104A 07NM-104B 07EV 104A  :

35C 07NM 1040 07SOV 1040 07NM 1040 07EV 104C

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35D 07NM 104D 07SOV 104B l 07NM-104D 07EV 104B .

Table 3.71, page 209  :

Delete the last sentence in Note 14. The sentence to be deleted reads as follows:

"The guide tube and valve assemblies are designated 07NM 104 A, B, C, and D for [

penetrations 35A, B, C, and D, respectively." {

Table 4.7 2, page 211 f

1. Delete penetration 35A. .

L 2. Penetration 35B Revise the valve number from *07NM-104B* to *07EV 104A."

Table 4.7 2, page 212

1. Penetration 350 Revise the valve number from *07NM 104C" to *07EV 104C."
2. Penetration 35D Revise the valve number from *07NM-104D* to *07EV 104B.*

II. PURPOSE OF THE PROPOSED CHANGES The Traversing incore Probe (TIP) system is used to calibrate the Local Power Range Monitors +

g (LPRMs). Durin0 the 1990 refueling outage the existing system will be replaced with a new state-of iho-art system. The new system will improve overall TIP system reliability, availability, and L accuracy.

L I

l Attachment 11 ,

. SAFETY EVALUATION REVISION 1 l Page 2 of 3 The modification replaces the four existing TIP drive units with three drive units. The helical shield cabling will be replaced with a titanium shield cable. Changing from a four channel system to a three channel system requires the use of one less contalnrnent penetration. As a result, i penetration X 35A will be capped. The titanium shielding, which does not require a nitrogen l i b'anket, permits the removal of the TIP system's nitrogen purge equipment. Penetration X 35E, which accommodated the TIP nitrogen purge line, will also be capped. 6 In addition, editorial changes were also made to correctly identify the containment isolation valve associated with Penetrations 358, C, and D.

1 111. IMPACT OF THE PROPOSED CHANGES The proposed changes revise Tables 3.71 and 4.7 2 to delete two penetrations, X 35A and X 35E, which will be capped as a result of the TIP system modification. The capping of these penetrations is not a safety concern because the replacement of an active isolation valve with a j passive barrier is a more reliab!e containment isolation method. The capped TIP penetrations will l be subjected to a Type A integrated leak rate test with visual inspection prior to plant restart and  ;

periodic testing in accordance with the requirements of 10 CFR 50, Appendix J. I

)

The changes do not alter the conclusions of the plant's accident analysos as documented in the FSAR or the NRC staff's SER nor are they considered a significant hazard consideration.

1 IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the James A. FitzPatrick Nuclear Power Plant in accordance with this proposed amendment would not involve a significant hazards consideration, as defined in 10 CFR 50.92, '

sinco the proposed changes would not:

1

1. Involve a significant increase in the probability or consequences of an accident previously evaluated. The changes update Tables 3.71 and 4.7 2 in support of a plant modification which caps ponetrations X 35A and X 35E. These penetrations will be leak tcsted during the l integrated leak rate test of the containment in accordance with the requirements of 10 CFR 50, Appendix J. The capping of these penetrations do not impact the plant's accident analyses as documented in the FSAR or the NRC staff's SER.

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! 2. create the possibility of a new or different kind of accident from those previously evaluated. j l The capping of these two containment penetrations improves containment isolation i l because a passive barrier is less likely to fall than an active Isolation valve. These changes l l do not introduce any new active failure modes and can not create a new or different kind of accident, i

3. Involve a significant reduction in the margin of safety. The propcsed changes cap two penetrations which are no longer required by the TIP system. Thece changes do not reduce the margin of safoty, since a passive barrier is a more reliablo method of containment  ;

isolation ther an active isolation valve.

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Attachment 11

~

r SAFETY EVALUATION REVISION 1  :

Page 3 of 3 V. IMPLEMENTATION OF THE PROPOSED CHANGES

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Implementation of the proposed changes will not impact the Al. ARA Program at FitzPatrick, nor will the changes impact the environment.

VI. CONCLUSION  :

These changes, as proposed, do not constitute an unreviewed safety question as defined in 10 CFR 50.59. That is, they:

a. will not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report;
b. will not increase the possibility for an accident or malfunction of a different type from any evaluated previously in the safety analysis report;
c. Will not reduce the margin of safety as defined in the basis for any technical specification; and l_ d. Involves no significant hazards consideration, as delined in 10 CFR 50.92.

I Vll. REFERENCES i 1. James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis Report, Sections

! 5.2A.6 and 7.5.9.

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2. James A. FitzPatrick Nuclear Power Plant Safety Evaluation Report (SER), dated November 20,1972 and Supplements.

( 3. James A. FitzPatrick Nuclear Power Plant Traversing Incore Probe (TIP) System Upgrade Conceptual Design Package Modification Number F189 253, December 1989.

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