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{{#Wiki_filter:9 ATTACHMENT PROPOSED CHANGE PRESSURE-TEMPERATURE RELATIONSHIP APPENDIX A TO PROVISIONAL OPERATING LICENSE DPR-16 OYSTER CREEK NUCLEAR GENERATING STATION GPU NUCLEAR CORPORATION DOCKET NO. 50-219 P
 
3.3  REACTOR COOLANT Applicability: Applies to the operating status of the reactor coolant system.
Obiective:      To assure the stmeture integrity of the reactor coolant system.
Specification: A. Pressure Temperature Relationships I
(i) Reactor Vessel Pressure Tests - the minimum reactor vessel temperature at a given pressure shall be in excess of that indicated by the curve A in Figures 3.3.1,3.3.2 and 3.3.3 for reactor operations to '
22, 27 and 32 effective full power years, respectively. The maximum
;                                      temperature for Reactor Vessel Pressure Testing is 250 F.
;                              (ii) Heatup and Cooldown Operations: Reactor noncritical -- the minimum reactor vessel temperature for heatup and cooldown j                                      operations at a given pressure when the reactor is not critical shall be in excess of that indicated by the curve B in Figures 3.3.1, 3.3.2 and
,                                      3.3.3 for reactor operations up to 22,27 and 32 effective full power years, respectively.
(iii) Power operations -- the minimum reactor vessel temperature for power operations at a given pressure shall be in excess of that indicated by the curve C in Figure 3.3.1,3.3.2 and 3.3.3 for reactor operations up to 22,27 and 32 effective full power years respectively.
Note: Curves A, B and C in Figures 3.3.1,3.3.2 and 3.3.3 apply when the closure head is on the reactor vessel and studs are fully tensioned.
(iv) Appropriate new pressure temperature limits must be generated when the reactor system has reached thirty two (32) effective full power years of reactor operation.
B. Reactor Vessel Closure Head Boltdown: The reactor vessel closure head studs may be elongated .020" (1/3 design preload) with no restrictions on reactor vessel temperature as long as the reactor vessel is at atmospheric pressure. Full tensioning of the studs is not permitted unless the temperature of the reactor vessel flange and closure head flange is in excess of 85 F.
C. Thermal Transients
: 1. The average rate of reactor coolant temperature change during normal heatup and cooldown shall not exceed 100 F in any one hour period.
: 2. The pump in an idle recirculation loop shall not be started unless the temperature of the coolant within the idle recirculation loop is within 50 F of the reactor coolant temperature.
OYSTER CREEK                                        3.3-1                Amendment No: 42,120,151
 
Transformation temperature. The minimum temperature for pressurization at any time in life has to account for the toughness propcrties in the most limiting regions of the reactor vessel, as well as the effects of fast neutron embrittlement.
Curves A, B and C on Figures 3.3.1,3.3.2 and 3.3.3 are derived from an evaluation of the      !
fracture toughness properties performed on the specimens coritained in Reactor Vessel          :
Materials Surveillance Program Capsule No. 2 (Reference 14). The results of dosimeter          !
wire analyses (Reference 14) indicated that the neutron fluence (E>l.0 MeV) at the end of 2
32 effective full power years of operation is 2.36 x 10" n/cm at the 1/4T (T= vessel wall thickness) location. This value was used in the calculation of the adjusted reference          !
nil-ductility temperature which, in turn, was used to generate the pressure-temperature        i curves A, B and C on Figures 3.3.1,3.3.2 and 3.3.3 (Reference 15). The 250 F maximum pressure test temperature provides ample margin against violation of the minimum              i required temperature. Secondary containment is not jeopardized by a steam leak during          ,
pressure testing, and the Standby Gas Treatment system is adequate to prevent unfiltered release to the stack.
l li Stud tensioning is considered significant from the standpoint of brittle fracture only when    f the preload exceed approximately 1/3 of the final design value. No vessel or closure stud      j minimum temperature requirements are considered necessary for preload values below 1/3 of the design preload with the vessel depressurized since preloads below 1/3 of the design      i preload result in vessel closure and average bolt stresses which are less than 20% of the      1 yield strengths of the vessel and bolting materials. Extensive service experience with these  ;
materials has confirmed that the probability of brittle fracture is extremely remote at these low stress levels, inespective of the metal temperature.
l The reactor vessel head flange and the vessel flange in combination with the double "O"        l ring type seal are designed to provide a leak tight seal when bolted together. When the        !
vessel head is placed on the reactor vessel, only that portion of the head flange near the      i inside of the vessel rests on the vessel flange. As the head bolts are replaced and tensioned, the vessel head is flexed slightly to bring together the entire contact surface adjacent to the "O" rings of the head and vessel flange. The original Code requirement was that boltup be done at qualification temperatures (T3OL) plus 60 F. Current Code requirements state (Ref.16) that for application of full bolt preload and reactor pressure up to 20% of hydrostatic test pressure, the RPV metal temperature must be at RTer or greater. The boltup temperature of 85 F was derived by determining the highest value of (T3OL + 60) and the highest value of RTmr, and by choosing the more conservative value of the two. Calculated values of(T3OL + 60) and RTmr of the RPV metal temperature were 85 F and 36 F, respectively (Ref.15). Therefore, selecting the boltup temperature to be 85 F provides 49 F margin over the current Code requirement based on          j RTmr.                                                                                          ;
Detailed stress analyses (4) were made on the reactor vessel for both steady state and transient conditions with respect to material fatigue. The results of these analyses are presented and compared to allowable stress limits in Reference (4). The specific conditions analyzed included 120 cycles of normal startup and shutdown with a heating and cooling rate of 100 F per hour applied continuously over a temperature range of            ;
100 F to 546 F and for 10 cycles of emergency cooldown at a rate of 300 F per hour            I applied over the same range. Thermal stresses from this analysis combined with the              I primary load OYSTER CREEK                              3.3-5                Amendment No: 15,42,120,151 j
 
==References:==
(1) FDSAR, Volume I, Section IV-2 (2) Letter to NRC dated May 19,1979, " Transient of May 2,1979"                          ,
(3) General Electric Co. Letter G-EN-9-55, " Revised Natural Circulation Flow Calculation", dated May 29,1979 (4) Licensing Application Amendment 16, Design Requirements Section (5) (Deleted)                                                                            i (6) FDSAR, Volume 1, Section IV-2.3.3 and Volume II, Appendix H
                                                                                                                      ~
                          -(7) FDSAR, Volume I, Table IV-2-1 (8) Licensing Application Amendment 34, Question 14 (9) Licensing Application Amendment 28, item III-B-2                                      l' (10) Licensing Application Amendment 32, Question 15 (11) (Deleted)                                                                            ,
(12) (Deleted)
(13) Licensing Application Amendment 16, Page 1                                          :
(14) GPUN TDR 725 Rev. 3: Testing and Evaluation of Irradiated Reactor Vessel Materials Surveillance Program Specimens (15) GENE-B13-01769 (GE Nuclear Energy): Pressure-Temperature Curves Per Regulatory Guide 1,99, Revision 2 for Oyster                        -
Creek Nuclear Generating Station.                  ,
(16) Paragraph G-2222(C), Appendix G, Section XI, ASME Boiler and Pressure Vessel Code,1989 Edition with 1989 Addenda,                        ;
                                      " Fracture Toughness Criteria for Protection Against Failure."
i l
l l
l l
OYSTER CREEK                          3.3-8a                    Amendment No: 135,140,151 I
 
1 FIGURE 3.3.1 l
OYSTER CREEK P-T CURVES VALID TO 22 EFPY l
1600                                                                                i i
                                              ^  _
22                                            ,
l swv 8        C 1400                                p    ,      7 l
l l
3 1200 G
l 0
l i
(
* 4
'        4.
o 1000                              ,
dm                                ;                x l                                                      I              CORE BELTUNE LIMITS I
E                                      j l        $    800                                    f                WITH ART OF 145'F FoR LOWER SHELL l        y                                                                PLATE G-8-6            j w
l g                          A - SYSTEM HYOROTEST LIMIT      I w                                                            WITH FUEL IN VESSEL        l l        !    600                  g                              B NON. NUCLEAR HEATUF/
[,              560 psig l
s                                      ;
f                    CootcowN UM:T e                                                        C NUCLEAR (CORE CRITICAU
        @                                                                    LIMIT 5                                Y 400 - n 5 Psio 1260F CURVES A,8,C ARE VAUO 200 -                  r soLTu,                                    FOR 22 EFPY OF OPERATION l
35'F MINIMUM CRIT 1CAUTY pr                          / TEMPERATURE = 96*F
                                      /
0              b/
0          100      200          300              400          500      600
;                              MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F) l l
l OYSTER CREEK                              3.3-9a                          Amendment NO. 151 l
l I
 
FIGURE 3.3 2 l                              OYSTER CREEK P-T CURVES VALID TO 27 EFPY 1600 A
re tarr      8  C 1400                                                                              I 3 1200                                ) l I  ,
O b
I
: n.                                                                                    ,
            ?                                                                                      j d
e
            $                                  r        r
            >                                                  1 w( CORE BELTLINE LIMITS E
            $    800                                        f          WITH ART OF 152'F FOR LOWER SHELL y                                                              PLATE G 8-6 w
2                                                      A . SYSTEM HYOROTEST Llwi
[                            WITH FUEL IN VESSEL E    600                                                B NON NUCLEAR p EATUS
[              560 psig                                      CooLDOWN *. MIT E
e                                                      C NUCLEAR (CCRE CRITICAL)
            $                                                                  UMIT 5                                )
400 -  27s PSio 0
126 F 200 -                      ;                    CURVES A.B.C ARE VALID soLTur          r sS*F FOR 27 EFPY OF OPERATION MINIMUM CNTICAUTY j/
                                                              ^
p                  TEMPERATURE = 96*F
                                          /
O O          100        200            300          400          500    600 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F) i l
3.3-9b OYSTER CREEK                                                                    Amendment No. W
 
                                ~
1 FIGURE 3.3,3 OYSTER CREEK P-T CURVES VALID TO 32 EFPY
                                                                                                            \
l 1600 A
n am B          C 1400 j        t 3 1200                                    /      i      !                          ,
I E
o                                                1      i b
I O
P-1000 w
w                                                          '
                                                                                                        ~
        >                                  f                                  CORE RELTLINE UMITS I
E                                                                      WITH ART OF 158'F ON                          )                  ;
FOR LOWER SHELL
                                                                                                        ~
PLATE G 8 6 W
g                                          J                    A SYSTEM HYOROTEST L'MIT g                              /                                                              _
W1TH FUEL IN ' ESSE.
D                            /
3    600                  /                                        B NON NUCLEAR HE ATUP.
        "                        g                                              CocLoewN uv:T        -
w          560 psig                                          i E
7                                                                C NUCLEAR (CCRE CRITICAU g                                                                            uMit            -
w E
j[
                                          ;      /                      CURVES A.B,C ARE VAUD
                  " "5 N                                                                            ~
o                                    FOR 32 EFPY OF OPERATION 126 F            t%
APP G REQUIREMhMT BASED -
ON CURVE A 91100 PSIG 200 - saTuP            r                                                              -
85*F
[/f d
_ _ _ MIMMUM CRITICAUTY TEMPERATURE = 96*F O                                                            t      t t,    f 0        100          200            300              400          500        600 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F) 3.3-9c OYSTER CREEK                                                                        Amendment No. 151
 
  ~
4.3 REACTOR COOLANT Applicability: Applies to the surveillance requirements for the reactor coolant system.
Objective:    To determine the condition of the reactor coolant system and the operation of the safety devices related to it.
Specification: A. Materials surveillance specimens and neutron flux monitors shall be installed in the reactor vessel adjacent to the wall at the midplane of the active core. Specimens and monitors shall be periodically removed, tested, and evaluated to determine the effects of neutron fluence on the fracture toughness of the vessel shell materials. The results of these evaluations shall be used to assess the adequacy of the P-T curves A, B and C in Figures 3.3.1,3.3.2 and 3.3.3. New curves shall be generated      l as required.
B. Inservice inspection of ASME Code Class 1, Class 2 and Class 3                I systems and components shall be performed in accordance with                  :
Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a(g)(6)(i).
C. Inservice testing cf ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a(g)(6)(i).
D. A visual examination for leaks shall be made with the reactor coolant        l system at pressure during each scheduled refueling outage or aller            i major repairs have been made to the reactor coolant system in accordance with Article 5000, Section XI. The requirements of specification 3.3.A shall be met during the test.                            ;
i E. Each replacement safety valve or valve that has been repaired shall be        I tested in accordance with subsection IWV-3510 of Section XI of the ASME Boiler and Pressure Vessel Code. Setpoints shall be as follows:
Number of Valves            Set Points (psig) 4                  1212i12 5                  1221    12 F. A sample of reactor coolant shall be analyzed at least every 72 hours for the purpose of determining the content of chloride ion and to check the conductivity.
OYSTER CREEK                            4.3-1        Amendment No.: 82,90,120,150,151,164 l
l
:}}

Latest revision as of 11:25, 13 July 2020

Proposed Tech Specs,Proposing New pressure-temp Limits Up to 22,27 & 32 EFPY Based on Predicted Nilductility Adjusted Ref Temp for Corrresponding EFPY of Operation
ML20117E706
Person / Time
Site: Oyster Creek
Issue date: 08/23/1996
From:
GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20117E700 List:
References
NUDOCS 9609030043
Download: ML20117E706 (9)
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Text

9 ATTACHMENT PROPOSED CHANGE PRESSURE-TEMPERATURE RELATIONSHIP APPENDIX A TO PROVISIONAL OPERATING LICENSE DPR-16 OYSTER CREEK NUCLEAR GENERATING STATION GPU NUCLEAR CORPORATION DOCKET NO. 50-219 P

3.3 REACTOR COOLANT Applicability: Applies to the operating status of the reactor coolant system.

Obiective: To assure the stmeture integrity of the reactor coolant system.

Specification: A. Pressure Temperature Relationships I

(i) Reactor Vessel Pressure Tests - the minimum reactor vessel temperature at a given pressure shall be in excess of that indicated by the curve A in Figures 3.3.1,3.3.2 and 3.3.3 for reactor operations to '

22, 27 and 32 effective full power years, respectively. The maximum

temperature for Reactor Vessel Pressure Testing is 250 F.
(ii) Heatup and Cooldown Operations
Reactor noncritical -- the minimum reactor vessel temperature for heatup and cooldown j operations at a given pressure when the reactor is not critical shall be in excess of that indicated by the curve B in Figures 3.3.1, 3.3.2 and

, 3.3.3 for reactor operations up to 22,27 and 32 effective full power years, respectively.

(iii) Power operations -- the minimum reactor vessel temperature for power operations at a given pressure shall be in excess of that indicated by the curve C in Figure 3.3.1,3.3.2 and 3.3.3 for reactor operations up to 22,27 and 32 effective full power years respectively.

Note: Curves A, B and C in Figures 3.3.1,3.3.2 and 3.3.3 apply when the closure head is on the reactor vessel and studs are fully tensioned.

(iv) Appropriate new pressure temperature limits must be generated when the reactor system has reached thirty two (32) effective full power years of reactor operation.

B. Reactor Vessel Closure Head Boltdown: The reactor vessel closure head studs may be elongated .020" (1/3 design preload) with no restrictions on reactor vessel temperature as long as the reactor vessel is at atmospheric pressure. Full tensioning of the studs is not permitted unless the temperature of the reactor vessel flange and closure head flange is in excess of 85 F.

C. Thermal Transients

1. The average rate of reactor coolant temperature change during normal heatup and cooldown shall not exceed 100 F in any one hour period.
2. The pump in an idle recirculation loop shall not be started unless the temperature of the coolant within the idle recirculation loop is within 50 F of the reactor coolant temperature.

OYSTER CREEK 3.3-1 Amendment No: 42,120,151

Transformation temperature. The minimum temperature for pressurization at any time in life has to account for the toughness propcrties in the most limiting regions of the reactor vessel, as well as the effects of fast neutron embrittlement.

Curves A, B and C on Figures 3.3.1,3.3.2 and 3.3.3 are derived from an evaluation of the  !

fracture toughness properties performed on the specimens coritained in Reactor Vessel  :

Materials Surveillance Program Capsule No. 2 (Reference 14). The results of dosimeter  !

wire analyses (Reference 14) indicated that the neutron fluence (E>l.0 MeV) at the end of 2

32 effective full power years of operation is 2.36 x 10" n/cm at the 1/4T (T= vessel wall thickness) location. This value was used in the calculation of the adjusted reference  !

nil-ductility temperature which, in turn, was used to generate the pressure-temperature i curves A, B and C on Figures 3.3.1,3.3.2 and 3.3.3 (Reference 15). The 250 F maximum pressure test temperature provides ample margin against violation of the minimum i required temperature. Secondary containment is not jeopardized by a steam leak during ,

pressure testing, and the Standby Gas Treatment system is adequate to prevent unfiltered release to the stack.

l li Stud tensioning is considered significant from the standpoint of brittle fracture only when f the preload exceed approximately 1/3 of the final design value. No vessel or closure stud j minimum temperature requirements are considered necessary for preload values below 1/3 of the design preload with the vessel depressurized since preloads below 1/3 of the design i preload result in vessel closure and average bolt stresses which are less than 20% of the 1 yield strengths of the vessel and bolting materials. Extensive service experience with these  ;

materials has confirmed that the probability of brittle fracture is extremely remote at these low stress levels, inespective of the metal temperature.

l The reactor vessel head flange and the vessel flange in combination with the double "O" l ring type seal are designed to provide a leak tight seal when bolted together. When the  !

vessel head is placed on the reactor vessel, only that portion of the head flange near the i inside of the vessel rests on the vessel flange. As the head bolts are replaced and tensioned, the vessel head is flexed slightly to bring together the entire contact surface adjacent to the "O" rings of the head and vessel flange. The original Code requirement was that boltup be done at qualification temperatures (T3OL) plus 60 F. Current Code requirements state (Ref.16) that for application of full bolt preload and reactor pressure up to 20% of hydrostatic test pressure, the RPV metal temperature must be at RTer or greater. The boltup temperature of 85 F was derived by determining the highest value of (T3OL + 60) and the highest value of RTmr, and by choosing the more conservative value of the two. Calculated values of(T3OL + 60) and RTmr of the RPV metal temperature were 85 F and 36 F, respectively (Ref.15). Therefore, selecting the boltup temperature to be 85 F provides 49 F margin over the current Code requirement based on j RTmr.  ;

Detailed stress analyses (4) were made on the reactor vessel for both steady state and transient conditions with respect to material fatigue. The results of these analyses are presented and compared to allowable stress limits in Reference (4). The specific conditions analyzed included 120 cycles of normal startup and shutdown with a heating and cooling rate of 100 F per hour applied continuously over a temperature range of  ;

100 F to 546 F and for 10 cycles of emergency cooldown at a rate of 300 F per hour I applied over the same range. Thermal stresses from this analysis combined with the I primary load OYSTER CREEK 3.3-5 Amendment No: 15,42,120,151 j

References:

(1) FDSAR, Volume I, Section IV-2 (2) Letter to NRC dated May 19,1979, " Transient of May 2,1979" ,

(3) General Electric Co. Letter G-EN-9-55, " Revised Natural Circulation Flow Calculation", dated May 29,1979 (4) Licensing Application Amendment 16, Design Requirements Section (5) (Deleted) i (6) FDSAR, Volume 1, Section IV-2.3.3 and Volume II, Appendix H

~

-(7) FDSAR, Volume I, Table IV-2-1 (8) Licensing Application Amendment 34, Question 14 (9) Licensing Application Amendment 28, item III-B-2 l' (10) Licensing Application Amendment 32, Question 15 (11) (Deleted) ,

(12) (Deleted)

(13) Licensing Application Amendment 16, Page 1  :

(14) GPUN TDR 725 Rev. 3: Testing and Evaluation of Irradiated Reactor Vessel Materials Surveillance Program Specimens (15) GENE-B13-01769 (GE Nuclear Energy): Pressure-Temperature Curves Per Regulatory Guide 1,99, Revision 2 for Oyster -

Creek Nuclear Generating Station. ,

(16) Paragraph G-2222(C), Appendix G,Section XI, ASME Boiler and Pressure Vessel Code,1989 Edition with 1989 Addenda,  ;

" Fracture Toughness Criteria for Protection Against Failure."

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OYSTER CREEK 3.3-8a Amendment No: 135,140,151 I

1 FIGURE 3.3.1 l

OYSTER CREEK P-T CURVES VALID TO 22 EFPY l

1600 i i

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22 ,

l swv 8 C 1400 p , 7 l

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3 1200 G

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(

  • 4

' 4.

o 1000 ,

dm  ; x l I CORE BELTUNE LIMITS I

E j l $ 800 f WITH ART OF 145'F FoR LOWER SHELL l y PLATE G-8-6 j w

l g A - SYSTEM HYOROTEST LIMIT I w WITH FUEL IN VESSEL l l  ! 600 g B NON. NUCLEAR HEATUF/

[, 560 psig l

s  ;

f CootcowN UM:T e C NUCLEAR (CORE CRITICAU

@ LIMIT 5 Y 400 - n 5 Psio 1260F CURVES A,8,C ARE VAUO 200 - r soLTu, FOR 22 EFPY OF OPERATION l

35'F MINIMUM CRIT 1CAUTY pr / TEMPERATURE = 96*F

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0 b/

0 100 200 300 400 500 600

MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F) l l

l OYSTER CREEK 3.3-9a Amendment NO. 151 l

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FIGURE 3.3 2 l OYSTER CREEK P-T CURVES VALID TO 27 EFPY 1600 A

re tarr 8 C 1400 I 3 1200 ) l I ,

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$ r r

> 1 w( CORE BELTLINE LIMITS E

$ 800 f WITH ART OF 152'F FOR LOWER SHELL y PLATE G 8-6 w

2 A . SYSTEM HYOROTEST Llwi

[ WITH FUEL IN VESSEL E 600 B NON NUCLEAR p EATUS

[ 560 psig CooLDOWN *. MIT E

e C NUCLEAR (CCRE CRITICAL)

$ UMIT 5 )

400 - 27s PSio 0

126 F 200 -  ; CURVES A.B.C ARE VALID soLTur r sS*F FOR 27 EFPY OF OPERATION MINIMUM CNTICAUTY j/

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p TEMPERATURE = 96*F

/

O O 100 200 300 400 500 600 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F) i l

3.3-9b OYSTER CREEK Amendment No. W

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1 FIGURE 3.3,3 OYSTER CREEK P-T CURVES VALID TO 32 EFPY

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l 1600 A

n am B C 1400 j t 3 1200 / i  ! ,

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P-1000 w

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FOR LOWER SHELL

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PLATE G 8 6 W

g J A SYSTEM HYOROTEST L'MIT g / _

W1TH FUEL IN ' ESSE.

D /

3 600 / B NON NUCLEAR HE ATUP.

" g CocLoewN uv:T -

w 560 psig i E

7 C NUCLEAR (CCRE CRITICAU g uMit -

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j[

/ CURVES A.B,C ARE VAUD

" "5 N ~

o FOR 32 EFPY OF OPERATION 126 F t%

APP G REQUIREMhMT BASED -

ON CURVE A 91100 PSIG 200 - saTuP r -

85*F

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_ _ _ MIMMUM CRITICAUTY TEMPERATURE = 96*F O t t t, f 0 100 200 300 400 500 600 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F) 3.3-9c OYSTER CREEK Amendment No. 151

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4.3 REACTOR COOLANT Applicability: Applies to the surveillance requirements for the reactor coolant system.

Objective: To determine the condition of the reactor coolant system and the operation of the safety devices related to it.

Specification: A. Materials surveillance specimens and neutron flux monitors shall be installed in the reactor vessel adjacent to the wall at the midplane of the active core. Specimens and monitors shall be periodically removed, tested, and evaluated to determine the effects of neutron fluence on the fracture toughness of the vessel shell materials. The results of these evaluations shall be used to assess the adequacy of the P-T curves A, B and C in Figures 3.3.1,3.3.2 and 3.3.3. New curves shall be generated l as required.

B. Inservice inspection of ASME Code Class 1, Class 2 and Class 3 I systems and components shall be performed in accordance with  :

Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a(g)(6)(i).

C. Inservice testing cf ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a(g)(6)(i).

D. A visual examination for leaks shall be made with the reactor coolant l system at pressure during each scheduled refueling outage or aller i major repairs have been made to the reactor coolant system in accordance with Article 5000,Section XI. The requirements of specification 3.3.A shall be met during the test.  ;

i E. Each replacement safety valve or valve that has been repaired shall be I tested in accordance with subsection IWV-3510 of Section XI of the ASME Boiler and Pressure Vessel Code. Setpoints shall be as follows:

Number of Valves Set Points (psig) 4 1212i12 5 1221 12 F. A sample of reactor coolant shall be analyzed at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the purpose of determining the content of chloride ion and to check the conductivity.

OYSTER CREEK 4.3-1 Amendment No.: 82,90,120,150,151,164 l

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