ML20094Q923
| ML20094Q923 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/28/1995 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Public Service Electric & Gas Co, Atlantic City Electric Co |
| Shared Package | |
| ML20094Q926 | List: |
| References | |
| RTR-REGGD-01.099, GL-88-011, NPF-57-A-088 NUDOCS 9512040186 | |
| Download: ML20094Q923 (11) | |
Text
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gg nro 21 UNITED STATES
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j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 20566 4001 PUBLIC SERVICE ELECTRIC & GAS COMPANY elLANTIC CITY ELECTRIC COMPANY DOCKET No 50-354 e
HOPE CREEK GEME@ TING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 88 License No. NPF-57 1.
The Nuclear Regulatory Comission (the Comission or the NRC) has found that:
A.
The application for amer.dment filed by the Public Service Electric &
Gas Company (PSE&G) dated July 27, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's roles 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 Comission; C.
There is reasonable assurance:
(i) that the activities authorized by this amendment can be conducted without endangering the health l
and safety of the public, and (ii) that such activities will be l
conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendent, and paragraph 2.C.(2) of Facility Operating License No. NPF-57 is hereby amended to read as follows:
(2) Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 88, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into the license.
)
PSE&G shall operate the facility in accordance with the Technical j
Specifications and the Environmental Protection Plan.
9512040186 951128 PDR ADOCK 05000354-p PDR
. 3.
The license amendment is effective as of its date of issuance, and shall be implemented within 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
_f
(
J n F. Stolz,'Directo ject Directorate I-ivision of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
November 28, 1995 t
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ATTACHMENT TO LICENSE AMENDMENT NO. 88 FACILITY OPERATING LICENSE NO. NPF-57 l
DOCKET NO. 50-354 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert i
3/4 4-21 3/4 4-21 3/4 4-22 3/4 4-22 3/4 4-23 3/4 4-23 3/4 4-23a 3/4 4-23b B 3/4 4-5 8 3/4 4-5 B 3/4 4-6 B 3/4 4-6 B 3/4 4-7 8 3/4 4-7 1
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REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM
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LIMITING CONDITION FOR OPERATION 3.4.6.1 The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1 (hydrostatic or leak testing), and Figure 3.4.6.1-2 (heatup by non-nuclear means, cooldown j
following a nuclear shutdown and low. power PHYSICS TESTS), and Figure 3.4.6.1-3 (operations with a critical core other than low power PHYSICS TESTS), withe a.
A maximum heatup of 100*F in any one hour period,
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b.
A maximum cooldown of 100*F in any one incur period, l
c.
A maximum temperature change of less than or equal to 20*F in any i
one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and
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d.
The reactor vessel flange and head flange metal temperature shall be mainta,ined greater than or equal to 79'F when reactor vessel head bolting studs are under tension.
l APPLICAEILITY: At all times.
1 ACTION:
Nith any of the above limits exceeded, restore the temperature and/or pressure to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the reactor coolant systems determine that the reactor coolant system remains acceptable for continued operations or be in at least HOT SHUTDONN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDONN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, t
l SURVEILLANCE REQUIRENENTS j
4.4.6.1.1 During system heatup, cooldown and inservice leak and hydrostatic l
testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3 l
as applicable, at least once per 30 minutes.
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RIACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be determined to be to the right of the criticality limit line of Figure 3.4.6.1-3 within 15 minutes prior to the withdrawal of control rods to bring l
the reactor to criticality and at least once per 30 minutes during system heatup.
4.4.6.1.3 The reactor vessel material surveillance specimens shall be removed and examined, to determine changes in reactor pressure vessel material properties, as required by 10 CFR 50, Appendix H.
The results of these examinations shall be used to update the curves of Figures 3.4.6.1-1,
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3.4.6.1-2, and 3.4.6.1-3.
4.4.6.1.4 The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to 70'F a.
In OPERATIONAL CONDITION 4 when reactor coolant system temperature is:
1.
s 100'F, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2.
s 80'F, at least once per 30 minutes.
- b. Within 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.
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' REACTOR COOLANT SYSTEM BASES 1/4.4.6 PRESBURE/ TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.
These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section (3.9) of the UFSAR.
During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
The operating limit curves of Figures 3.4.6.1-1, 3.4.6.1-2, and 3,4.6.1-3 are derived from the fracture toughness requirements of 10 CFR 50 Appendix G and ASME Code Section III, Appendix G.
The curves are based on the RTmn and stress intensity factor information for the reactor vessel components.
Fracture toughness limits and the basis for compliance are more 3
fully discussed in UFSAR Chapter 5, Paragraph 5.3.1.5,
" Fracture Toughness."
l The reactor vessel materials have been tested to determine their initial RTmn.
The results of some of these, tests are shown in Table B 3/4.4.6-1.
Reactor operation and resultant fast neutron, E greater than 1 MeV, irradiation will cause an increase in the RTmn.
Therefore, an adjusted reference temperature, based upon the fluence, nickel content and copper content of the material in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommendations of Regulatory Guide 1.99, Rev. 2,
" Radiation Embrittlement of Reactor Vessel Material".
The pressure /
temperature limit curves, Figures 3.4.6.1-1, 3,4.6.1-2, and 3.4.6.1-3, includes an assumed shift in RTmg for the end of life fluence.
The actual shift in RTmn of the vessel material will be established periodically during operation by removing a.nd evaluating, irradiated flux wires installed near the inside wall of the reactor vessel in the core area, since the neutron spectra at the flux wires and vessel inside radius are essentially identical, the irradiated flux wires can be used with confidence in predicting reactor vessel material transition temperature shift. The operating limit curves of Figures 3.4.6.1-1, 3.4.6.1-2, and 3.4.6.1-3 shall be adjusted, as required, on the basis of the flux wire data and recommendations of Regulatory Guide 1.99, Rev. 2.
REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)
The pressure-temperature limit lines shown in Figures 3.4.6.1-1 and 3.4.6.1-3, curves for inservice leak and hydrostatic testing and reactor criticality have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.
The number of reactor vessel irradiation surveillance capsules and the frequencies for removing and testing the specimens in these capsules are provided in UFSAR Section 5.3 and Appendix 5A.
3/4.4.7 MAIN STRAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break. Only one valve in each line is required to maintain the integrity of the containment, however, single failure considerations require that two valves be OPERABLE. The surveillance requirements are based on the operating history of this type valve. The maximum closure time has been selected to contain fission products and to ensure the cor6 is not uncovered following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.
3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.
Components of the reactor coolant system were designed to provide access
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to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1977 Edition and Addenda through Summer 1978.
The inservice inspection program for ASME Code Class 1, 2 and 3 components will be performed in accordance with Section XI of the ASME Boiler j
and Pressure Vessel Code and applicable addenda as required by 10 CFR Part j
50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR Part 50.55a(g) (6) (i).
)
3/4.4.9 RESIDUAL HEAT REMOVAL 1
A single shutdown cooling mode loop provides sufficient heat removal l
capability for removing core decay heat and mixing to assure accurate temperature indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and the.t an alternate method of coolant mixing be in operation.
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BASES TABLE B 3/4.4.6-1 REACTOR VESSEL TOUGHNESS HEAT / SLAB PREDICTED EOL BELTLINE WELD SEAM I.D.
OR HIGHEST UPPER SHELF MAX. BOL COMPOIEENT OR MAT'L TYPE HEAT / LOT CU(%)
Mi(%)
RT g (
- F)
ART.eg (
- F)
(FT-LBS)
ELorM Plate SA-533 GR B CL.1 5K3025-1
.15 0.71
+19 53.5 67 72.5 Weld vert. seems for D53040/
.08 0.59
-30 G.8 120 32-8 shells 4E5 1125-02205 NOTE:
- These values are given only for the benefit of calculating the end-of-life (EOL) RTrer
HEAT / LOT Rl.or I M Shell Ring Connected to SA 533, GR.B, C1.1 All Heats.
+19 Vessel Flange Botton Head Dome SA 533, GR.B. C1.1 All Heats
+30 Bottom Head Torus SA 533, GR.B. C1.1 All Heats
+30 LPCI Nozzles" SA 508, C3.2 All Heats
-20 Top Head Torus SA 533, GR.B. C1.1 All Heats
+19 Top Head Flange SA 508, C1.2 All Heats
+10 Vessel Flange SA 508, C1.2 All Heats
+10 Feedwater Nozzle SA 508, C1.2 All Heats
-20 Weld Metal All RPV Welds All Heats O
Closure Stude SA 540, GR.B. 24 All Heats Meet 45 ft-lbs & 25 alls lateral expansion at +10*F (1) The design of the Hope Creek vessel results in these nozzles experiencing a predicted ROL fluence at
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1/4T of the vessel thickness of 2.81 x 10" n/cm. Therefore, these nozzles are predicted to have an ROL l
2 1
l RT,eg of +24*F.
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