ML17285B335
| ML17285B335 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 06/14/1990 |
| From: | Larkins J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17285B336 | List: |
| References | |
| NUDOCS 9006220074 | |
| Download: ML17285B335 (17) | |
Text
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~kg**4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 WASHINGTON PUBLIC POWER SUPPLY SYSTEM DOCKET NO. 50-397 NUCLEAR PROJECT NO.
2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 87 License No. NPF-21 1.
The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for amendment filed by the Washington Public Power Supply System (the licensee),
dated October 27, 1989 as supplemented by letter dated April 5, 1990 comply with the standards and requirements of the Atomic Energy Act of
- 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B.
C.
D.
E.
The facility will operate in conformity with the applica-tion, the provisions of the Act, and the regulations of the Commission; There is reasonable assurance (i) that the activities author-ized 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 Commission's regula-tions set forth in 10 CFR Chapter I; 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 W
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.
9006220074 9006i4 PDR ADOCK 05000397 P
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No.
NPF-21 is hereby amended to read as follows:
(2)
Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.
87, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION John T. Larkins, Acting Director Project Directorate V
Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specif ications Date of Issuance:
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ATTACHMENT TO LICENSE AMENDMENT NO.
FACILITY OPERATING LICENSE NO. NPF-21 DOCKET NO. 50-397 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
Also to be replaced are the following overleaf pages.
AMENDMENT PAGE xx(a) 3/4 4-18 3/4 4-19 3/4 4-20 3/4 4-21 B3/4 4-4 B3/4 4-5 B3/4 4-6 OVERLEAF PAGE 3/4 4-17 3/4 4-22 B3/4 4-3
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LIST OF FIGURES INDEX FIGURE 3.2.4-5 LINEAR HEAT GENERATION RATE (LHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE GE11 LEAD FUEL ASSEMBLIES ~
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PAGE 3/4 2"lOD 3.2. 6-1 3.2. 7-1 3.2. 8-1 3.4.1. 1"1 3.4.6.1 4.7-1
- 3. 9; 7-1 B 3/4 3-1 B 3/4.4.6"1
- 5. 1-1
- 5. 1-2 5.1-3 OPERATING REGION LIMITS OF SPEC. 3.2.6...............
3/4 2-12 OPERATING REGION LIMITS OF SPEC. 3.2.7...............
3/4 2-14 OPERATING REGION LIMITS OF SPEC. 3.2.8........... -...
3/4 2-16 THERMAL POWER LIMITS OF SPEC. 3.4.1.1-1..............
3/4 4-3a MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE.......................
3/4 4-20 SAMPLE PLAN 2)
FOR SNUBBER FUNCTIONAL TEST..........
3/4 7-15 HEIGHT ABOVE SFP WATER LEVEL VS.
MAXIMUM LOAD TO BE CARRIED OVER SFP.....................................
3/4 9-10 FAST NEUTRON FLUENCE (E>1MeV) AT 1/4 T AS A FUNCTION OF SERVICE LIFE.............................
EXCLUSION AREA BOUNDARY.............................
LOW POPULATION ZONE..................................
B 3/4 4"7 5-2 5"3 UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIOACTIVE GASEOUS AND LIQUID EFFLUENTS.............
5-4 REACTOR VESSEL WATER LEVEL...........................
B 3/4 3"8 WASHINGTON NUCLEAR - UNIT 2 xx(a)
Amendment No.
.8~
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TABLE 4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT AND ANALYSIS 1.
Gross Beta and Gamma Activity Determination 2.
Isotopic Analysis for 00SE E(UIVALENT I-131 Concentration 3.
Radiochemical for K Determination 4.
Isotopic Analysis for Iodine 5.
Isotopic Analysis of an Off-gas Sample Including quantitative Measurements for at least Xe-133, Xe-135 and Kr-88 SAMPLE AND ANALYSIS FEEI UEEEIEIUUY At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> At least once per 31 days At least once per 6 months*
a)
At least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever the specific activity exceeds a limit, as required by ACTION b.
b)
At least one sample, between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following the change.in THERMAL POWER or off-gas level, as required by ACTION c.
At least once per 31 days OPERATIONAL CONDITIONS IN WHICH SAMPLE AND ANALYSIS RE UIRED 1, 2, 3
lf, 28, 3k, 4F 1,
2 "Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.
PUntil the specific activity of the primary coolant system is restored to within its limits.
REACTOR COOLANT SYSTEM 3/4.4. 6 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM 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) curve A for hydrostatic or leak testing; (2) curve B for heatup by non-nuclear
- means, cooldown following a nuclear shutdown and low power PHYSICS TESTS; and (3) curve C for operations with a critical core other than low power PHYSICS TESTS, with:
a.
A maximum heatup of 100 F in any 1-hour period, b.
A maximum cooldown of 100 F in any 1-hour period, c.
A maximum temperature change of less than or equal to 204F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and d.
The reactor vessel flange and head flange temperature greater than or equal to 804F when reactor vessel head bolting studs are under tension.
APPLICABILITY: At all times.
ACTION:
With 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 system; determine that the reactor coolant system remains acceptable for continued operations or be in at least HOT SHUTDOWN within 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 />.
SURVEILLANCE RE UIREMENTS 4.4.6.1.1 During system heatup,
- cooldown, and inservice leak and hydrostatic 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 Figure 3.4.6.1 curves A, 8 or C, as applicable, at least once per 30 minutes.
WASHINGTON NUCLEAR - UNIT 2 3/4 4-18 Amendment No.
87
REAL'TOR COOLANT SYSTEM SURVEILLANCE RE UIREMENTS Continued) 4.4.6.1.2 The reactor coolant system temperature and pressure shall be deter-mined to be to the right of the criticality limit line of Figure 3.4.6. 1 curve C within 15 minutes prior to the withdrawal of control rods to bring 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 Part 50, Appendix H in accordance with the schedule in Table 4.4.6.1.3-1.
The results of these examinations shall be used to update the curves of Figure 3.4. 6.1.
4.4.6.1.4 The reactor vessel flange and head flange temperature shall be verified to be greater than or equal to 80'F:
a.
In OPERATIONAL CONDITION 4 when reactor coolant system temperature 15:
I.
< 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.
< 90'F, at least once per 30 minutes.
b.
Mithin 30 minutes prior to and at least once per 30 minutes during tensioning of the reactor vessel head bolting studs.
MASHINGTON NUCLEAR - UNIT 2 3/4 4"19 Amendment No. 87
A B
C CORE.BELTLINE LIMITS lE0 o I
%9
'0 K9 A, B, C, - ABC LIMITS AFTER AN ASSUMED 75 F
CORE*
BELTLINE TEMP SHIFT FROM AN INITIAL RTNDT OF 28 F
110 F
312 PSIG 180 F
312 PSIG 140 F
312 PSIG BOLTUP LIMIT 80 F 0
50 100 150 200 250 XO XO 400 450 500 MINIMUM REACTOR VESSEL METAL TEMPERATURE Temperature F
FIGURE 3.4.6.1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VERSUS REACTOR VESSEL PRESSURE WASHINGTON NUCLEAR " UNIT 2 3/4 4-20 Amendment No.
87
THIS PAGE IS BLANK INTENTIONALLY WASHINGTON NUCLEAR - UNIT 2 3/4 4-21 Amendment No.
TASLE 4.4.6. 1. 3"1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM-WITHDRAWAL SCHEDULE CAPSULE NUBBER VESSEL LOCATION 300 120 304 LEAD FACTOR Due to symmetry, all capsules are expected to have the same lead factor.
LF = 1.2 at the 1/4T LF = 0.86 at vessel ID WITHDRAWAL TIME EFPY Standby
REACTOR COOLANT SYSTEM BASES 3/4.4. 5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body doses resulting from a main steam line failure outside the containment during steady-state operation will not exceed small fractions of the dose guidelines of 10 CFR Part 100.
The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.
These values are conserva-tive in that specific site parameters, such as SITE BOUNDARY location and neteorologfcal condftions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limfted time periods wfth the primary coolant's specific activity greater than 0.2 ricrocurfe per gram DOSE EQUIVALENT I-131, but less than or equal to 4.0 micro-curies per gram DOSE E/UIVALENT I-131, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.
Closing the eafn steam line isolation valves prevents the release of activity to the envfrons should a steam line rupture occur outside containment.
The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected fn sufficient time to take corrective action.
MASHINGTON NUCLEAR - UN?T 2 8 3/4 4-3 Amendment No.
39
REACTOR COOLANT SYSTEM BASES 3/4.4.6 PRESSURE/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 4.9 of the FSAR.
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.
During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.
These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure.
Therefore, a pressure-temperature curve based on steady-state conditions, i.e.,
no thermal stresses, represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.
The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the control-ling location.
The thermal gradients established during heatup produce tensile stresses which are already present.
The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.
Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling
- location, each heatup rate of interest must be analyzed on an individual basis.
The reactor vessel materials have been tested to determine their initial RTNDT The results of these tests are shown in Table B 3/4.4.6-1.
Reactor operation and resultant fast neutron irradiation, E greater than I MeV, will cause an increase in the RTNDT.
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, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials."
The pressure/
temperature limit curve, Figure 3.4.6.1 includes predicted adjustments for this shift in RTNDT for the end of life fluence and is effective for 10 EFPY.
The actual shift in RTN>T of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73 and 10 CFR Part 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area.
The irradiated specimens can be used with confidence in predicting reactor ves-sel material tran'sition temperature shift.
The operating limit curves of Figure 3.4.6. 1 shall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99, Revision 2.
MASHINGTON NUCLEAR - UNIT 2 B 3/4 4-4 Amendment No.
BASES
)
~ )h PRESSURE/TEMPERATURE LIMITS (Continued)
The pressure-temperature'imit lines shown in Figure 3.4.6.1 for reactor criticality and for inservice leak and hydrostatic testing 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.
3/4.4.7 MAIN STEAM 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 core 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.
Access to permit inservice inspections of components of the reactor coolant system is in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1974 Edition and Addenda through Summer 1975.
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 and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR 50.55a(g)(6)(i).
3/4.4.9 RESIDUAL HEAT REMOVAL A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate tempera-ture indication, however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation.
MASHINGTON NUCLEAR - UNIT 2 B 3/4 4-5 Amendment No. 8~
BASES TABLE B 3/4.4.6-1 REACTOR VESSEL TOUGHNESS MATERIAL TYPE COMPONENT BELTLINE Ring 1 Plate SA-533, GRB, CL1 Ring 2 Plate SA-533,
- GRB, CL1 Girthweld E8018NM Girthweld RAC01NMM NON"8ELTLINE Ring 3 Plate SA-533, GRB, CLl Ring 4 Plate SA-533,
- GRB, CL1 Plate Top Head Side Plates Bottom Head Dollar SA-533,
- GRB, CL1 Plates Bottom Head Radial Plates Nozzles SA-508, CL2 Flange Bolt Studs SA-540, B23 "Regulatory Guide 1.99, Revision 2, calculated hRTNDT
>100
>100
+28 41 33 36 15 0.15 0.6
-10 0.15 0.5
-30 0.03 1.01 N.A.
0.08 0.8 N.A.
-50 "44 SA-533, GRB, CLl SA"533,
- GRB, CL1 HIGHEST MAXIMUM STARTING 50 FT"LB/35 h
MIN. UPPER SHELF CU Ni RT MIL TEMP F RT FT"LB L
F
~ll F