ML20085F535
| ML20085F535 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 03/09/1995 |
| From: | Mark Miller NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20085F538 | List: |
| References | |
| NUDOCS 9506190291 | |
| Download: ML20085F535 (25) | |
Text
_
t' o.
I g&P EICg,
[.
+
., a UNITED STATES j
NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 20046 4 001
%.... /
I PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE 1
Amendment No. 98 License No. DPR-80 j
1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Pacific Gas and Electric Company (the licensee) dated August 17, 1994, complies 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.
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 conducted in compliance with the Commission's regulations; 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 i
requirements have been satisfied.
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-80 is hereby amended to read as follows:
l l
l 9506190291 950309 PDR ADOCK 05000275 P
PDR A
s' 4
- (2)
Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.
98, are hereby incorporated in the license.
Pacific Gas and Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
3.
This license amendment is effective as of 90 days from the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION O,
1 Melanie A. Miller, Senior Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 9, 1995
s' o**C'cg[*
UNITED STATES yo i.
'j NUCLEAR REGULATORY COMMISSION
- k WASHINGTON, D.C. *am aaat oq-.....,f PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-323 DIABLO CANYON NUCLEAR POWER PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE l
Amendment No. 97 License No. DPR-82 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Pacific Gas and Electric Company (the licensee) dated August 17, 1994, complies 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.
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 Commission's regulations; 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.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-82 is hereby amended to read as follows:
I i
d'
. (2)
Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.
97, are hereby incorporated in the license.
Pacific Gas and Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.
3.
This license amendment is effective as of 90 days from the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
'YL k ay A Melanie A. Miller, Senior Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 9, 1995 I
l 1
1
l e
ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 98 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO. 97 TO FACILITY OPERATING LICENSE NO. OPR-82 DOCKET NOS. 50-275 AND 50-321 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed.nages. The revised pages are identified by the captioned amendment number and ontain marginal lines indicating the area of change. Overleaf pages are also included, as appropriate.
REMOVE INSERT vii vii viii viii xv xv 3/4 4-7 3/4 4-7 3/4 4-8 3/4 4-8 3/4 4-22 3/4 4-22 3/4 4-23 3/4 4-24 3/4 4-34 3/4 4-34 3/4 4-37 3/4 4-37 3/4 4-38 3/4 4-38 B 3/4 4-5 B 3/4 4-5 B 3/4 4-7 B 3/4 4-7 B 3/4 4-15 B 3/4 4-15 B 3/4 4-16 B 3/4 4-16 6-15c 6-15c
INDEX 4
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PAGI 3/4.3 INSTRUMENTATION (continued)
Chl ori ne Detecti on Systems...............................
3/4 3-54 Explosive Gas Monitoring Instrumentation.................
3/4 3-59 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................
3/4 3-60 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation..............................
H o t S t a n d by..............................................
3 / 4 4 - 1 3/4 4-2 Hot Shutdown.............................................
3/4 4-3 Cold Shutdown - Loops Filled.............................
3/4 4-5 Cold Shutdown - Loops Not F111ed.........................
3/4 4-6 3/4.4.2 SAFETY VALVES l
0perating................................................
3/4 4-8 3/4.4.3 PRESSURIZER..............................................
3/4 4-9 3/4.4.4 RELIEF VALVES...........................................
3/4 4-10 3/4.4.5 STEAM GENERATORS........................................
3/4 4-11 DIABLO CANYON - UNITS 1 & 2 vii Amendment Nos. 57 1 00, 70 a 74, 98 &
ItiDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION EAGE 3/4.4 REACTOR COOLANT SYSTEM (continued)
TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION................
3/4 4-16 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION.......................
3/4 4-17 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.............................
3/4-4-18 Operational Leakage...................................
3/4 4-19 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES......
3/4 4-21 1
3/4.4.8 SPECIFIC ACTIVITY.....................................
3/4 4-25 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY
> 1 pCI/ GRAM DOSE EQUIVALENT I-131....................
3/4 4-27 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..................................
3/4 4-28 3/4.4.9 PRESSURE / TEMPERATURE LIMITS..........................
3/4 4-30 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -
APPLICABLE UP TO 8 EFPY..............................
3/4 4-31 FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -
APPLICABLE UP TO 8 EFPY..............................
3/4 4-32 l
Overpressure Protection Systems......................
3/4 4-35 DIABLO CANYON - UNITS 1 & 2 vili Amendment Nos. 54 a 53,98 8 ;
i
g BASES SECTION PEi.E 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE..........................
B 3/4 4-3 g
3/4.4.8 SPECIFIC ACTIVITY.......................................
B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS.............................
B 3/4 4-7 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS............................................
B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS.............................
B 3/4 5-1 3/4.5.4 BORON INJECTION SYSTEM..................................
B 3/4 5-2 3/4.5.5 REFUELING WATER STORAGE TANK............................
B 3/4 5-3 DIABLO CANYON - UNITS 1 & 2 xv Amendment Nos. 54 a 5; 98 &
itfDil
~
BASES SECTION EAjf 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 CONTAINMENT..............................................
B 3/4 6-1 3:1.6.2 DEPRESSURIZATION AND COOLING SYSTEMS.....................
B 3/4 6-3 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................
B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L..................................
B 3/4 6-4 3 /4.7 PLANT SYSTEMS 3/4.7.I TURBINE CYCLE............................................
B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION..........
B 3/4 7-3 3/4.7.3 VITAL COMPONENT COOLING WATER SYSTEM.....................
5 3/4 7-3 3/4.7.4 AUXILIARY SALTWATER SYSTEM...............................
B 3/4 7-3 3/4.7.5 CONTROL ROOM VENTILATION SYSTEM..........................
B 3/4 7-3 3/4.7.6 AUXILIARY BUILDING SAFEGUARDS AIR FILTRATION SYSTEM......
B 3/4 7-4 3/4.7.7 SNUBBERS.................................................
B 3/4 7-4 3/4.7.B SEALED SOURCE CONTAMINATION..............................
B 3/4 7-6 3/4.7.11 AREA TEMPERATURE MONITORING..............................
B 3/4 7-7 3/4.7.12 ULTIMATE HEAT SINK.......................................
B 3/4 7-7 3/4.7.13 FLOOD PROTECTION.........................................
B 3/4 7-8 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and DNSITE POWER DISTRIBUTION...............................
B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES..................
B 3/4 8-3 DIABLO CANYON - UNITS 1 & 2 xvi Amendment Nos.75 & 74
(
This Page Intentionally Deleted DIABLO CANYON - UNITS 1 & 2 3/4 4-7 Amendment Nos. 98 & 97
o' l
REACTOR COOLANT SYSTEM OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer Code safety valves shall be OPERABLE with a lift setting of 2485 psig i 1%.*
APPLICABILITY: MODES 2, 3 and 4.#
ACTION:
a.
With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HOT SHUTDOWN with all RCS cold leg temperatures less than or equal to 323*F within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
The provisions of Specification 3.0.4 may be suspended for up to 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> per valve for entry into and during operationr in MODE 3 and 4#
l for the purpose of setting the pressurizer Code sdety valves under ambient (hot) conditions provided a preliminary cold setting was made prior to heatup.
SURVEILLANCE RE0VIREMENTS 4.4.2.2 No additional requirements other than those required by Specifica-tion 4.0.5.
- The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
- When all RCS cold leg temperatures are greater than 323*F.
DIABLO CANYON - UNITS 1 & 2 3/4 4-8 Amendment Nos. 49 5 iS, 72 1 71,98 & !
r-TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION 1.
8948 A, B, C, and D Accumulator,lRHRandSIS first off check valves from RCS cold legs 2.
8819 A, B, C, and D SIS second off check valves from RCS cold legs 3.
8818 A, B, C, and D RHR second off check valves from RCS cold legs 4.
8956 A, B, C, and D Accumulator second off check valves from RCS cold legs 5.
B701* and 8702*
RHR suction isolation valves 6.
8949# A, B, C, and D RHR and SIS first off check valves from RCS hot legs 7.
8905# A, B, C, and D SIS second off check valves from RCS hot legs 8.
8740# A and B RHR second off check valves from RCS hot legs 9.
8802*# A and B SIS to RCS hot legs isolation valves 10.
8703*#
RHR to RCS hot legs isolation valve i
- Testing per Specification 4.4.6.2.2c. not required.
- For flowpaths with 3 pressure isolation valves in series, at least 2 of the 3 valves shall meet the requirements of Specification 3.4.6.2f.
DIABLO CANYON - UNITS 1 & 2 3/4 4-21 Amendment Nos. 16 and 15 l
_j
\\
i i
This Page Intentionally Deleted I
e 9
1 DIABLO CANYON - UNITS 1 & 2 3/4 4-22 Amendment Nos. 98 & 97 (next page 3/4 4-25)
1 1:
h li i,
P THIS PAGE INTENTIONALLY DELETED.
l 1
l l
DIABLO CANYON - UNITS 1 & 2 3/4 4-33 Amendment Nos. 54 & 53
1 4
This Page Intentionally Deleted DIABLO CANYON - UNITS I & 2 3/4 4-34 Amendment Nos.98 & 97
e 9
This Page Intentionally Deleted i
DIABLO CANYON - UNITS 1 & 2 3/4 4-37 Amendment Nos. 55 a 54, 98 &
l l
i t
i i
i l
i f
This Page Ir.+entionally Deleted t
i i
i i
DIABLO CANYON - UNITS 1 & 2 3/4 4-38 Amendment Nos. 98 & 97
=-
REACTOR COOLANT SYSTEM BASES 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appro-priately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam generator leakage rate of I gps. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conserva-tive in that specific site parameters of the Diablo Canyon site, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity greater than 1 microcurie / gram DOSE EQUIVALENT I-131, but within the a'llowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phen >menon which may occur following changes in THERMAL POWER. Operation with spe:ific activity levels exceeding 1.0 microcurie / gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-1 should be limited since the activity levels allowed by figure 3.4-1 increase the 2-hour thyroid dose at tiie 5fTE BOUNDARY by a factor of up to 20 following a postulated steam generator tube rupture.
DIABLO CANYON - UNITS 1 & 2 B 3/4 4-5 Amendment Nos.10 a IfT 98 & ' i
BASES i
SPECIFIC ACTIVITY (Continued)
The sample analysis for determining the gross specific activity and I can exclude the radioiodines because of the low reactor coolant limit of 1 microcurie /
gram DOSE EQUIVALENT I-131, and because, if the limit is exceeded, the radio-iodine level is to be determined every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
If the gross specific activity level and radioiodine level in the reactor coolant were at their limits, the radioiodine contribution would be approximately 1%.
In a release of reactor l
coolant with a typical mixture of radioactivity, the actual radioiodine contri-bution would probably be about 20%. The exclusion of radionuclides with half-lives less than 10 minutes from these determinations has been made for several The first consideration is the difficulty to identify short-lived reasons.
radionuclides in a sample that requires a significant time to collect, trans-port, and analyze.
The second consideration is the predictable delay time between the postulated release of radioactivity from the reactor coolant to its release to the environment and transport to the SITE BOUNDARY, which-is relat-able to at least 30 minutes decay time. The choice of 10 minutes for the half-life cutoff was made because of the nuclear characteristics of the typical reactor coolant radioactivity.
The radionuclides in the typical reactor coolant I
have half-lives of less than 4 minutes or half-lives of greater than 14 minutes, which allows a distinction between the radionuclides above and below a half-life of 10 minutes.
For these reasons the radionuclides that are excluded from con-i sideration are expected to decay to very low levels before they could be trans-ported from the reactor coolant to the SITE BOUNDARY under any accident condi-tion.
Based upon the above considerations for excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> between sample taking and completing the initial analysis is based upon a typical time necessary to per-form the sampling, transport the sample, and perform the analysis of about 90 minutes.
After 90 minutes, the gross count should be made in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties.
The counter should be reset to a reproducible efficiency versus energy.
It is not necessary to identify specific nuclides.
The radiochemical determination of nuclides should be based on multiple counting of the sample within typical counting basis following sampling of less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, about 1 day, about I week, and about 1 month.
Alternatively, gamma spectroscopy may be used.
Reducing T,yg to less than 500*F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospher's steam relief valves.
The Surveillance Requirements provide adequate asst-.ce that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.
Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena.
A reduction in fre-quency of isotopic analyses following power changes may be permissible if justified by the data obtained.
DIABLO CANYON - UNITS I & 2 B 3/4 4-6
REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Presure Vessel Code,Section III, Appendix G:
1.
The reactor coolant temperature and pressure and system heatup and cool-down rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the service period specified thereon:
a.
Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown.
Limit lines for cooldown rates between those presented may be obtained by interpolation; and b.
Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only.
For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
2.
These limit lines shall be calculated periodically using methods provided
- below, 3.
The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F, 4.
Deleted 5.
System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI. Allowable pressures and tempera-tures for inservice leak and hydrostatic tests are given in Figure 3.4-2.
6.
The criticality limit on Figure 3.4-2 is based on the minimum allowable temperature of 295'F for an inservice hydrostatic test of 110% of operating pressure.
The fracture toughness testing of the ferritic materials in the reactor vessel was performed in accordance with the 1966 Edition for Unit I and the 1968 Edition for Unit 2 of the ASME Boiler and Pressure Vessel Code, Sec-tion III.
These properties are then evaluated in accordance with the NRC Standard Review Plan.
Heatup and cooldown limit curves are calculated using the most limiting value of the nil ductility reference temperature, RT at the end of 8 effective full power years (EFPY) of service life. Ne, 8 EFPY service life period is chosen such that the limiting RT, at the 1/4T location in the core region DIABLO CANYON - UNITS I & 2 B 3/4 4-7 Amendment Nos. 64-t-W; 98 & 9
A 6
h THIS PAGE INTENTIONALLY DELETED.
l I
e l
I DIABLD CANYON - UNITS 1 & 2 3 3/4 4-3 Amendment Nos. 54 and 53
-~ e
4 REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp.
Furthermore, since the thermal stresses at the outside are tensile and increase with increasing heatup rate, a lower bound curve cannot be defined.
Rather, each heatup rate of interest must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be 'the lesser of the three values taken from the curves under consideration.
The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.
I LOW TEMPERATURE OVERPRESSURE PROTECTION 1
The OPERABILITY of both Class 1 PORVs or an RCS vent opening of at least 2.07 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 323*F.
Either Class 1 PORV has adequate relieving capability to protect the RCS from overpressurization for all anticipated transients.
~
i DIABLO CANYON - UNITS 1 & 2 8 3/4 4-15 Amendment Nos. 31 & S0, 98 &
,-~ --
e REACTOR COOLANT SYSTEM BASES LOW TEMPERATURE OVERPRESSURE PROTECTION (Continued)
The Maximum Allowed PORY Setpoint for the LTOPs will be modified, if required, based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in Table 4.4-5.
l l
l l
DIABLO CANYON - UNITS 1 & 2 B 3/4 4-16 Amendment Nos. 98 & 97 l
_--------------------__--------_------------------_--J
ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 6)
Limitations on the operability and use of the liquid and gaseous affluent treatment system; to ensure that the appro-priate portions of these systems are used to reduce releases
)
of radioactivity when the projected doses in a 31 -day 1
period would exceed 2 percent of the guidelines for the annual dose or dose caenitment conforming to Appendix I to t
t 7)
Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to i
areas at or beyond the SITE B0UNDARY shall be finited to the following:
i n.
For noble gases: Less than or equal to a dose rate of 500 mrom/yr to the whole body and less thar. or equal to a dose rate of 3000 ares /yr to the skin, and i
b.
For Iodine-131, for Iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 ares /yr to any organ.
8)
Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous affluents from each unit to areas beyond the SITE SOUNDARY conforming to Appendix I to 10 CFR part 50, 9)
Limitatiens on the annual and quarterly doses to MEMBERS OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to j
areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR part 50, and Limitations on the annual dose or dose comitment to any 10 MEMBER OF THE puBLIC due to releases of radioactivity and to j
radiation from uranium fuel cycle sources conforming to 4 CFR part 190.
h.
Radiolootcal Environmental Monitorino Procram A program shall be provided to monitor the radiation and radio-nuclides in the environs of the plant. The program shall provide i
(1) representative measurements of radioactivity in the highest potential exposure pathways, and of the affluent monitoring program (2) verification of the accuracy and modeling of environmental exposure pathways. The program shall (1) be contained in the i
RMCP, (2) conform to the gu' dance of Appendix ! to 10 CFR part 50, and (3) include the following:
)
1)
Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ERMP, i
DIABLO CANYON - UNITS 1 & 2 6-15b Amendment Nos. 85 8 84
f ADMINISTRATIVE CONTROLS t
l t
PROCEDURES AND PROGRAMS (Continued) h.
Radioloaical Environmental Monitorina Proaram l
2)
A Land Use Census to ensure that changes in the use of areas at l
and beyond the SITE BOUNDARY are identified and that modifica-tions to the monitoring program are made if required by the results of this census, and 3)
Participation in a Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in the environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.
1.
Reactor Coolant Puno Flywheel Insoection Inspect each reactor coolant pump flywheel in accordance with the recommendations of Regulatory Position c.4.b of Regulatory Guide 1.14, Revision 1, August 1975.
I DIABLO CANYON - UNITS I & 2 6-15c Amendment Nos. 05 1 04, 98 f
..