ML20136D591

Amend 188 to License DPR-16,updating pressure-temperature Limits Up to 22,27 & 32 Effective Full Power Years

ML20136D591
Person / Time
Site:	Oyster Creek
Issue date:	03/06/1997
From:	Milano P NRC (Affiliation Not Assigned)
To:
Shared Package
ML20136D595	List: ML20136D591 ML20136D600
References
NUDOCS 9703120411
Download: ML20136D591 (11)

Text

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l Y Elly UNITED STATES y

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NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 20056 0001

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GPU NUCLEAR CORPORATION AllD JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET NO. 50-219 1

OYSTER CREEK NUCLEAR GENERATING STATION 1

AMENDMENT TO FACILITY OPERATING LICENSE-Amendment No.188 License No. DPR-16 1.

The Nuclear Regulatory Commission (the Comission) has found that:

4 A.

The application for amendment by GPU Nuclear Corporation, et al.,

(the licensee) dated August 23, 1996, as supplemented January 8, 1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application. the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; i

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

i 9703120411 970306 ADOCK0500g9 PDR P

i i I 2.

Accordingly, the license is amended by changes to the Technical i

Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-16 is hereby amended to read as follows:

(')

Technical Snecifiqations 2

The Technical Specifications contained in Appendices A and B, as revised through Amendment No.188, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of issuance, to be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMISSION 0.

Patrick D. Milano, Acting Director Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: March 6,1997 i

I

l ATTACHMENT TO LICENSE AMENDMENT N0. 188 FACILITY OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 Replace the following pages of the Appendix A, Technical Specifications, with the attached pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remova Insert 3.3-1 3.3-1 3.3-5 3.3-5 3.3-8a 3.3-8a 3.3-9 3.3-9a 3.3-9a 3.3-9b 3.3-9b 3.3-9c 3.3-9c 4.3-1 4.3-1 4.3-2 4.3-2

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3.3 REACTOR COOLANT Aeolicability:

Applies to the operating status of the reactor coolant system.

Obiective:

To assure the structure integrity of the reactor coolant system.

Specification:

A. Pressure Temocrature Relationshios (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 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 inFigures 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 ofreactor operation.

B. Reactor Vessel Closure Head Bqhda_wn 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.

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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 j

50 F of the reactor coolant temperature.

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

188

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. Transformation trmperature. The minimum temperature for pressurization at any time in life has to account for the toughness properties in the most limiting regions of the reactor

~ vessel, as well as the effects of fast neutron embrittlement.

i 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 contained 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>1.0 MeV) at the end of l

32 effective full power years of operation is 2.36 x 10" n/cm at the 1/4T (T= vessel wall 2

l 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 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 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.

Stud tensioning is considered significant from the standpoint of brittle fracture only when the preload exceed approximately 1/3 of the final design value. No vessel or closure stud

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minimum temperature requirements are considered necessary for preload values below 1/3 l

of the design preload with the vessel depressurized since preloads below 1/3 of the design preload result in vessel closure and average bolt stresses which are less than 20% of the 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, irrespective of the metal temperature.

The reactor vessel head flange and the vessel flange in combination with the double "O" 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 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 donc 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 RTer, and by choosing the more conservative value of the two. Calculated values of(T3OL + 60) and RTer f the RPV metal o

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 rte 1 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 i

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 applied over the same range. Thermal stresses from this analysis combined with the primary load OYSTER CREEK 3.3-5 Amendment No: 15,42,120,151.

188 l

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

(3) General Electric Co. Letter G EN 9-55, " Revised Natural Circulation Flow Calculation", dated May 29,1979 (4) Licenthg Application Amendment 16, Design l

Requirements Section (5) (Deleted) i (6) FDSAR, Volume 1, Section IV-2.3.3 and Volume 11, Appendix H (7) FDSAR, Volume 1, Table IV-2-1 (8) Licensing Application Amendment 34, Question 14 (9) Licensing Application Amendment 28, item lil-B-2 (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 l

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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4 OYSTER CREEK-3.3-8a Amendment No: 135,140,151, 188 i

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l FIGURE 3.3.1 l

l OYSTER CREEK P-T CURVES VALID TO 22 EFPY l

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C. NUCLEAR (CORE CRITICAU f

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400 " 37s Psio 1260F CURVES A,8,C ARE VAUD 200 -- SoLTue '

r FOR 22 EFPY OF OPERATION -

Ss'F MINIMUM CRITICAUTY

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1. TEMPanATURE - se'r p

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

OYSTER CREEK 3.3-9a Amendment NO. 151, 188

i FIGURE 3.3.2 OYSTER CREEK P-T CURVES VALID TO 27 EFPY 1000 A

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A. SYSTEM HYOMC*f $? LtW*

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FOR 27 EFPY OF OPERATION es*F M198MUhl CRITICAUTV s

r p-TEMPERATURE = es*F j

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0 100 200 300 400 500 000 MenMUM REACTOR VESSEL METAL TEMPERATURE (*F)

,3.3-9b OYSTER CREEK Amendment No. 151.

188

FIGURE 3.3.3 OYSTER CREEK P-T CURVES VALID TO 32 EFPY

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FOR 32 EFPY OF OPERATION 3

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% APP G REQUMEMENT SASED ON CURVE A $ 1100 PSIG 200 -

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3.3-9c OYSTER CREEK Amendment No. 151' N

.4.3 REACTOR COOLANT Applicability:

Applies to the surveillance requirements for the reactor coolant system.

Obiective:

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

Materials surveillance specimens and neutron flux monitors shall be Soecification:

A.

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 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(gX6)(i).

C.

Insersice testing of 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 system at pressure during each scheduled refueling outage or after 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.

E.

Each replacement safety valve or valve that has been repaired shall be 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 (osin) 4 1212112 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,.188

i.

G.

Primary Coolant System Pressure Isolation Valves Soccification:

l.

Periodic leakage testing "' on each valve listed in Table 4.3.1 shall be accomplished prior to exceeding 600 psig reactor pressure every time the plant is placed in the cold shutdown condition for refueling, each time the plant is placed in a cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in the preceding 9 months, whenever the valve is moved whether by manual actuation or due to flow conditions, and aAer returning the valve to service mRer maintenance, repair or replacement work is performed.

H.

Reactor Coolant System Leakane 1.

Unidentified leakage rate shall be calculated at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

2.

Total leakage rate (identified and unidentified) shall be calculated at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

3.

A channel calibration of the primary containment sump flow integrator and the primary containment equipment drain tank flow integrator shall be conducted at least once per 18 months.

1.

An inservice inspection program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the NRC staff positions on schedule, methods, personnel, and sample expansion included in the generic letter or in accordance with alternate measures approved by the NRC staff.

Bases Data is available relating neutron fluence (E>1.0MeV) and the change in the Reference Nil-Ductility Transition Temperature (RTun). The pressure-temperature (P-T) operating curves A, B, and C in Figures 3.3.1,3.3.2, and 3.3.3 were developed based on the results of testing and l

evaluation of specimens removed from the vessel after 8.38 EFPY. of operation. Similar testing and analysis will be performed throughout vessel life to monitor the effects of neutron irradiation on the reactor vessel shell materials.

The inspection program will reveal problem areas should they occur, before a leak develops. In addition, extensive visual inspection for leaks will be made on critical systems. Oyster Creek was designed and constructed prior to

• To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

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• NRC Order dated April 20,1981.

OYSTER CREEK 4.3-2 Amendment No.: 82,90,97,118,120,151,154, 188

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