ML050900399

From kanterella
Jump to navigation Jump to search

Technical Specification Pages Re Alternative Source Term
ML050900399
Person / Time
Site: Vermont Yankee Entergy icon.png
Issue date: 03/29/2005
From:
Office of Nuclear Reactor Regulation
To:
References
TAC MC0253
Download: ML050900399 (14)
v • d • e


Text

VYNPS 1.0 DEFINITIONS Z. Surveillance Interval - Relocated to Specification 4.0.1.

AA. Deleted BB. Source Check - The qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

CC. Dose Equivalent I-131 - The dose equivalent I-131 shall be that concentration of I-131 (microcurie/gram) which alone would produce the same dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present. The dose conversion factors used for this calculation shall be those listed in Federal Guidance Report (FGR) 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," 1988; FGR 12, "External Exposure to Radionuclides In Air, Water, and Soil," 1993; or NRC Regulatory Guide 1.109, Revision 1, October 1977.

DD. Deleted EE. Deleted FF. Deleted GG. Deleted HH. Deleted II. Deleted JJ. Deleted KK. Deleted LL. Deleted MM. Deleted NN. Core Operating Limits Report - The Core Operating Limits Report is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.6.C.

Plant operation within these operating limits is addressed in individual specifications.

Amendment No. 4A, QL3, 44, -.G, 84, 4-03, 6.,- 154-1,4-6&, 4-74, 4-&8, 4-94, 2-21, 223 5

I VYNPS BASES:

3.2 PROTECTIVE INSTRUMENTATION In addition to reactor protection instrumentation which initiates a reactor scram, station protective instrumentation 5as been provided which initiates action to mitigate the consequences of accidents which are beyond the reactor operator's ability to control, or terminate a single operator error before it results in serious consequences. This set of Specifications provides the limiting conditions of operation for the primary system isolation function and initiation of the core standby cooling and standby gas treatment systems. The objectives of the Specifications are (i) to assure the effectiveness of any component of such systems even during periods when portions of such systems are out of service for maintenance, testing, or calibration, and (ii) to prescribe the trip settings required to assure adequate performance. This set of Specifications also provides the limiting conditions of operation for the control rod block system and surveillance instrumentation.

Isolation valves are installed in those lines that penetrate the primary containment and must be isolated during a loss-of-coolant accident so that the radiation dose limits are not exceeded during an accident condition.

Actuation of these valves is initiated by protective instrumentation shown in Table 3.2.2 which senses the conditions for which isolation is required.

Such instrumentation must be available whenever primary containment integrity is required. The objective is to isolate the primary containment so that the limits of 10CFR5O.67 are not exceeded during an accident. The objective of the low turbine condenser vacuum trip is to minimize the radioactive effluent releases to as low as practical in case of a main condenser failure. Subsequent releases would continue until operator action was taken to isolate the main condenser unless the main steam line isolation valves were closed automatically on low condenser vacuum. The manual bypass is required to permit initial startup of the reactor during low power operation.

The instrumentation which initiates primary system isolation is connected in a dual channel arrangement. Thus, the discussion given in the bases for Specification 3.1 is applicable here.

The low reactor water level instrumentation is set to trip when reactor water level is 127N above the top of the enriched fuel. This trip initiates closure of Group 2 and 3 primary containment isolation valves. For a trip setting of 127" above the top of the enriched fuel, the valves will be closed before perforation of the clad occurs even for the maximum break and, therefore, the setting is adequate.

The top of the enriched fuel (351.50 from vessel bottom) is designated as a common reference level for all reactor water level instrumentation. The intent is to minimize the potential for operator confusion which may result from different scale references.

The low-low reactor water level instrumentation is set to trip when reactor water level is 82.5" 1H20 indicated on the reactor water level instrumentation above the top of the enriched fuel. This trip initiates closure of the Group 1 primary containment isolation valves and also activates the ECCS and RCIC System and starts the standby diesel generator system. This trip setting level was chosen to be low enough to prevent spurious operation, but high enough to initiate ECCS operation and primary system isolation so that no melting of the fuel cladding will occur, and so that post-accident cooling can be accomplished and the limits of 10CFR50.67 will not be violated.

Amendment No. 69, aI, BV1, 0v 52, ;, 223 75

I VYNPS BASES: 3.2 (Cont'd)

For the complete circumferential break of 28-inch recirculation line and with the trip setting given above, ECCS initiation and primary system isolation are initiated in time to meet the above criteria. The instrumentation also covers the full range of spectrum breaks and meets the above criteria.

The high drywell pressure instrumentation is a backup to the water level instrumentation, and in addition to initiating ECCS, it causes isolation of Group 2, 3, and 4 isolation valves. For the complete circumferential break discussed above, this instrumentation will initiate ECCS operation at about the same time as the low-low water level instrumentation, thus, the results given above are applicable here also. Certain isolation valves including the TIP blocking valves, CAD inlet and outlet, drywell vent, purge and sump valves are isolated on high drywell pressure. However, since high drywell pressure could occur as the result of non-safety-related causes, such as not venting the drywell during startup, complete system isolation is not desirable for these conditions and only certain valves are required to close. The water level instrumentation initiates protection for the full spectrum of loss of coolant accidents and causes a trip of certain primary system isolation valves.

Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident. In addition to monitoring steam flow, instrumentation is provided which causes a trip of Group 1 isolation valves.

The primary function of the instrumentation is to detect a break in the main steam line, thus only Group 1 valves are closed. For the worst case accident, main steam line break outside the drywell, this trip setting of 140 percent of rated steam flow in conjunction with the flow limiters and main steam line valve closure limit the mass inventory loss such that fuel is not uncovered, cladding temperatures remain less than 12950 F and release of radioactivity to the environs is well below 10CFR5O.67.

Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in this area. Trips are provided on this instrumentation and when exceeded cause closure of Group 1 isolation valves.

Its setting of ambient plus 95'F is low enough to detect leaks of the order of 5 to 10 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, it is a backup to high steam flow instrumentation discussed above, and for small breaks, with the resultant small release of radioactivity, gives isolation before the limits of 10CFR50.67 are exceeded.

Isolation of the condenser mechanical vacuum pump (MVP) is assumed in the safety analysis for the control rod drop accident (CRDA). The MVP isolation instrumentation initiates closure of the MVP suction isolation valve following events in which main steam line radiation monitors exceed a predetermined value. A High Main Steam Line Radiation Monitor trip setting for MVP isolation of 5 3 times background at rated thermal power (RTP) is as low as practicable without consideration of spurious trips from nitrogen-16 spikes, instrument instabilities and other operational occurrences.

Isolating the condenser MVP limits the release of fission products in the event of a CRDA.

Pressure instrumentation is provided which trips when main steam line pressure drops below 800 psig. A trip of this instrumentation results in closure of Group 1 isolation valves. In the refuel, shutdown, and startup modes, this trip function is provided when main steam line flow exceeds 40%

of rated capacity. This function is provided primarily to provide protection against a pressure regulator malfunction which would cause the Amendment No. a, a&, 84, 8, i, BI: 01 52, 212, 223 76

VYNPS BASES: 3.6 and 4.6 (Cont'd)

A Note is included in Figure 3.6.2 that specifies test instrumentation 0

uncertainty must be +/- 2 F and the flange region temperatures must be 0

maintained greater than or equal to 72 F when using such instrumentation in lieu of permanently installed instrumentation.

Qualified test instrumentation may only be used for the purpose of maintaining the temperature limit when the vessel is vented and the fluid level is below the flange region. If permanently installed instrumentation (with a 10*F uncertainty) is used during head tensioning and detensioning operations, the 80'F limit must be met.

In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures will be maintained within 500 F of each other prior to startup of an idle loop.

Vermont Yankee is a participant in the Boiling Water Reactor Vessel and Internals Project Integrated Surveillance Program (ISP). for monitoring changes in the fracture toughness properties of ferritic materials in the reactor pressure vessel (RPV) beltline region. (See UFSAR Section 4.2 for additional ISP details.) As ISP capsule test reports become available for RPV materials representative of VYNPS, the actual shift in the reference temperature for nil-ductility transition (RTMT) of the vessel material may be re-established. In accordance with Appendix H to 10CFR50, VY is required to review relevant test reports and make a determination of whether or not a change in Technical Specifications is required as a result of the surveillance data.

B. Coolant Chemistry A steady-state radioiodine concentration limit of 1.1 JCi of I-131 dose equivalent per gram of water in the Reactor Coolant System can be reached if the gross radioactivity in the gaseous effluents is near the limit, as set forth in the Offsite Dose Calculation Manual, or if there is a failure or prolonged shutdown of the cleanup demineralizer.

Limits on the maximum allowable level of radioactivity in the reactor coolant are established to ensure that in the event of a release of any radioactive material to the environment during a design basis accident, radiation doses are maintained within the limits of 10CFR50.67.

The Limiting Conditions for Operation contain iodine specific activity limits. The iodine isotopic activities per gram of reactor coolant are expressed in terms of DOSE EQUIVALENT 1-131. The allowable levels are intended to limit the 2-hour radiation dose to an individual at the site boundary to within 10CFR50.67 dose guidelines.

The iodine spike limit of four (4) microcuries of I-131 dose equivalent per gram of water provides an iodine peak or spike limit for the reactor coolant concentration to assure that the radiological consequences of a postulated LOCA are within 10CFR50.67 dose guidelines.

The reactor coolant sample will be used to assure that the limit of Specification 3.6.B.1 is not exceeded. The radioiodine concentration would not be expected to change rapidly during steady-state operation over a period of 96 hours4 days <br />0.571 weeks <br />0.132 months <br />. In addition, the trend of the radioactive gaseous effluents, which is continuously monitored, is a good indicator of the trend of the radioiodine concentration in the reactor coolant.

When a significant increase in radioactive gaseous effluents is indicated, as specified, an additional reactor coolant sample shall be taken and analyzed for radioactive iodine.

Amendment No. X}, 4X0, 9o, 6&, a6-, }4A, 243, -2, 223 140

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION at normal cooldown rates if the torus water temperature exceeds 120 0 F.

e. Minimum Water Volume

- 68,000 cubic feet

f. Maximum Water Volume

- 70,000 cubic feet

2. Primary containment 2. The primary containment integrity shall be integrity shall be maintained at all times demonstrated as required when the reactor is by the Primary critical or when the Containment Leakage Rate I reactor water temperature Testing Program (PCLRTP).

is above 2120 F and fuel is in the reactor vessel except while performing low power physics tests at atmospheric pressure at power levels not to exceed 5 Mw(t).

3. (Blank)
3. If a portion of a system that is considered to be an extension of primary containment is to be opened, isolate the affected penetration flow path by use of at least one closed and deactivated automatic valve, closed manual valve or blind flange.
4. In accordance with the
4. Whenever primary PCLRTP, verify that the containment integrity is following leakage rates required:

are within acceptable limits:

a. The leakage rate from any one main steam
a. The leakage rate isolation valve (MSIV) through each MSIV; shall not exceed 62 scfh at 44 psig (Pa);
b. The combined leakage rate for the main steam
b. The combined leakage pathways; and rate from the main steam pathways shall
c. The combined leakage not exceed 124 scfh at rate for the secondary 44 psig (Pa); and containment bypass pathways.
c. The combined leakage rate from the secondary containment bypass pathways shall not exceed 5 scfh at 44 psig (Pa).

Amendment No. 64, 52a, -6, ;4-223 147

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION

i. Suspend movement of irradiated fuel assemblies and the fuel cask in secondary containment; and ii. Suspend core alterations; and iii. Initiate action to suspend operations with the potential for draining the reactor vessel.

C. Secondary Containment System C. Secondary Containment System

1. Secondary Containment 1. Surveillance of secondary Integrity shall be containment shall be maintained during the performed as follows:

following modes or conditions: a. A preoperational secondary

a. Whenever the reactor containment is in the Run Mode, capability test Startup Mode, or Hot shall be conducted Shutdown condition*; after isolating the or Reactor Building and placing either Standby Gas Treatment System filter train in operation. Such tests shall demonstrate the capability to maintain a 0.15 inch of water vacuum under calm wind (2 < u < 5 mph) condition with a filter train flow rate of not more than 1550 cfm.

I

  • NOTE: The reactor mode switch may be changed to either the Run or Startup/Hot Standby position, and operation not considered to be in the Run Mode or Startup Mode, to allow testing of instrumentation associated with the reactor mode switch interlock functions, provided:

0

1. Reactor coolant temperature is < 212 F;
2. All control rods remain fully inserted in core cells containing one or more fuel assemblies; and
3. No core alterations are in progress.

-14, 147, 197, 223 155a Amendment No.

VYNPS 3.7 LIMITING CONDITIONS FOR 4.7 SURVEILLANCE REQUIREMENTS OPERATION

b. ,During movement of b. Additional tests irradiated fuel shall be performed assemblies or the during the first fuel cask in operating cycle secondary under an adequate number of different containment; or environmental wind conditions to enable
c. During alteration of valid extrapolation the Reactor Core; or of the test results.
d. During operations c. Secondary with the potential containment for draining the capability to reactor vessel. maintain a 0.15 inch of water vacuum under calm wind (2<U<5 mph) conditions with a filter train flow rate of not more than 1550 cfm, shall be demonstrated at I least quarterly and at each refueling outage prior to refueling.

Amendment No. 14, a-9-, 223 156

I VYNPS BASES:

3.7 STATION CONTAINMENT SYSTEMS A. Primary Containment The integrity of the primary containment and operation of the core standby cooling systems in combination limit the off-site doses to values less than to those suggested in 10CFR50.67 in the event of a break in the primary system piping. Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists. Concern about such a violation exists whenever the reactor is critical, above atmospheric pressure and temperature above 212 0F. An exception is made to this requirement during initial core loading and while a low power test program is being conducted and ready access to the reactor vessel is required.

The reactor may be taken critical during the period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring. Procedures and the Rod Worth Minimizer would limit control worth to less than 1.30% delta k.

The pressure suppression pool water provides the heat sink for the reactor primary system energy release following postulated rupture of the system. The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown for normal operating pressure.

Since all the gases in the drywell are purged into the pressure suppression chamber air space during a loss-of-coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid must not exceed 62 psig, the allowable internal design pressure for the pressure suppression chamber. The design volume of the suppression chamber (water and air) was obtained by considering that the total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber (Reference Section 5.2 FSAR).

Using the minimum or maximum water volumes given in the specification, the calculated peak accident containment pressure is approximately 44 psig, which is below the ASME design pressure of 56 3

psig. 3 ' The minimum volume of 68,000 ft results in a submergency of approximately four feet. The majority of the Bodega tests 2) were run with a submerged length of four feet and with complete condensation. Thus, with respect to downcomer submergence, this specification is adequate.

The maximum temperature at the end of blowdown tested during the Humbolt Bay(') and Bodega Bay tests was 170'F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperature above 170*F.

(1) Robbins, C. H., "Tests on a Full Scale 1/48 Segment of the Humbolt Bay Pressure Suppression Containment", GEAP-3596, November 17, 1960.

(2) Bodega Bay Preliminary Hazards Summary Report, Appendix 1, Docket 50-205, December 28, 1962.

(3) Internal design pressure is 62 psig.

Amendment No. Ltr did , y 01 52, Q-4,223 163

VYNPS BASES: 4.7 (Cont'd)

The primary containment preoperational test pressures are based upon the calculated primary containment pressure response in the event of a loss-of-coolant accident. The calculated peak accident containment pressure would be about 44 psig, which would reduce to 27 psig within about 20 seconds following the pipe break. The suppression chamber pressure rises to about 25 psig within 30 seconds, equalizes with drywell pressure, and then decays with drywell pressure.(')

The ASME design pressure of the drywell and absorption chamber is 3

56 psig. (2 The design leak rate is 0.5%/day at this pressure. As pointed out above, the pressure response of the drywell and suppression chamber following an accident would be the same after about 10 seconds. Based on the primary containment pressure response and the fact that the drywell and suppression chamber function as a unit, the primary containment will be tested as a unit rather than the individual components separately.

Maintaining the primary containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Primary Containment Leakage Rate Testing Program (PCLRTP) required by Specification 6.7.C. The PCLRTP specifies the leakage rate test requirements, schedules, and acceptance criteria for tests of the leak tight integrity of the primary reactor containment and systems and components which penetrate the containment.

The PCLRTP implements the leakage rate testing of the primary containment as required by 10CFR50.54(o) and 10CFR5O, Appendix J, Option B as modified by approved exemptions. The leakage limits prescribed by the PCLRTP are consistent with the design of VYNPS and the analytical models used to calculate the radiological consequences of design basis accidents described in the Updated Final Safety Analysis Report.

Consistent with the limiting assumptions used in the associated accident consequence analyses, the PCLRTP differentiates three leakage pathways to the environment: (1) primary containment leakage to secondary containment, which is filtered through the standby gas treatment system before being released via the plant stack; (2) main steam pathways; and (3) secondary containment bypass pathways. Leakage effluent from the main steam and secondary containment bypass pathways have different pathways (ground level) to the environment than the leakage into secondary containment. These pathways are defined in the PCLRTP.

(1) Section 14.6 of the FSAR.

(2) 62 psig is the maximum internal design pressure for this ASME design (56 psig) pressure.

167 Amendment No. 3-1, BZ 04-152, 22 3

I VYNPS BASES: 4.7 (Contsd)

The test frequencies are adequate to detect equipment deterioration prior to significant defects, but the tests are not frequent enough to load the filters, thus reducing their reserve capacity too quickly.

That the testing frequency is adequate to detect deterioration was demonstrated by the tests which showed no loss of filter efficiency after 2 years of operation in the rugged shipboard environment on the NS Savannah (ORNL 3726) . Pressure drop tests across filter sections are performed to detect gross plugging of the filter media.

Considering the relatively short time that the fans may be run for test purposes, plugging is unlikely, and the test interval is reasonable.

Such heater tests will be conducted once during each operating cycle.

Considering the simplicity of the heating circuit, the test frequency is sufficient. Air distribution tests will be conducted once during each operating cycle.

The in-place testing of charcoal filters is performed using a halogenated hydrocarbon, which is injected into the system upstream of the charcoal filters. Measurements of the challenge gas concentration upstream and downstream of the charcoal filters is made. The ratio of the inlet and outlet concentrations gives an overall indication of the leak tightness of the system. Although this is basically a leak test, since the filters have charcoal of known efficiency and holding capacity for. elemental iodine and/or methyl iodine, the test also gives an indication of the relative efficiency of the installed system.

High-efficiency particulate air filters are installed before and after the charcoal filter to minimize potential release of particulates to the environment and to prevent clogging of the iodine filters. An efficiency of 99% is adequate to retain particulates that may be released to the Reactor Building following an accident. This will be demonstrated by testing with DOP as testing medium.

The efficiencies of the particulate and charcoal filters are sufficient to prevent exceeding 10CFR5O.67 limits for the accidents analyzed. The analysis of post-accident hydrogen purge assumed a charcoal filter efficiency of 95%. Hence requiring in-place test efficiencies of 99t for these filters and a laboratory methyl iodide test of 97.5% for the charcoal provides adequate margin.

The test interval for filter efficiency was selected to minimize plugging of the filters. In addition, testing for methyl iodide removal efficiency will be demonstrated. This will be done either by removal of a charcoal sample cartridge which contains charcoal equivalent to the bed thickness or removing one adsorber tray from the system and using the charcoal therein, after mixing, to obtain at least two samples equivalent to the bed thickness. Any HEPA filters found defective should be replaced with filters qualified according to Regulatory Position C.3.d of Regulatory Guide 1.52. If laboratory test results are unacceptable, all charcoal adsorbent in the system should be replaced with charcoal adsorbent qualified according to Regulatory Guide 1.52.

Amendment No. 6, a449, 223 170

VYNPS BASES: 4.7 (Cont'd)

D. Primary Containment Isolation Valves Those large pipes comprising a portion of the reactor coolant system whose failure could result in uncovering the reactor core are supplied with automatic isolation valves (except those lines needed for emergency core cooling system operation or containment cooling) . The closure times specified herein and per Specification 4.6.E are adequate to prevent loss of more cooling from the circumferential rupture of any of these lines outside the containment than from a steam line rupture.

Therefore, the isolation valve closure times are sufficient to prevent uncovering the core.

Purge and vent valve testing performed by Allis-Chalmers has demonstrated that all butterfly purge and vent valves installed at Vermont Yankee can close from full open conditions at design basis containment pressure. However, as an additional conservative measure, limit stops have been added to valves 16-19-7/7A, limiting the opening of these valves to 500 open while operating, as requested by NRC in their letter of May 22, 1984. (NVY 84-108)

In order to assure that the doses that may result from a steam line break do not exceed the 10CFR50.67 guidelines, it is necessary that no fuel rod perforation resulting from the accident occur prior to closure of the main steam line isolation valves. Analyses indicate the fuel rod cladding perforations would be avoided for the main steam valve closure times, including instrument delay, as long as 10.5 seconds.

The test closure time limit of five seconds for these main steam isolation valves provides sufficient margin to assure that cladding perforations are avoided and 10CFR5O.67 limits are not exceeded.

Redundant valves in each line ensure that isolation will be effected applying the single failure criteria.

The main steam isolation valves are primary containment isolation valves and are tested in accordance with the requirements of the Inservice Testing program.

The containment is penetrated by a large number of small diameter instrument lines. The flow check valves in these lines are tested for operability in accordance with Specification 4.6.E.

E. Reactor Building Automatic Ventilation System Isolation Valves (RBAVSIVs)

In the event that there are one or more RBAVSIVs inoperable when secondary containment integrity is required, the affected penetrations that have been isolated must be verified to be isolated on a periodic basis. This is necessary to ensure that those penetrations required to be isolated following an accident, but no longer capable of being automatically isolated, will be in the isolated position should an event occur. The verification frequency of once per 31 days is appropriate because the valves are operated under administrative controls and the probability of their misalignment is low.

Verification of isolation does not require any testing or device manipulation. Rather, it involves verification that the affected penetration remains isolated.

The RBAVSIVs covered by this surveillance requirement, along with their test requirements, are included in the Inservice Testing Program.

Amendment No. 1, mZ&, a-8, 4-9, 223 171

VYNPS BASES:

3.8 RADIOACTIVE EFFLUENTS A. Deleted B. Deleted C. Deleted D. Liquid Holdup Tanks The tanks listed in this Specification include all outdoor tanks that contain radioactivity that are not surrounded by liners, dikes, or walls capable of holding the tank contents, or that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10CFR Part 20.1001-20.2402, Appendix B, Table 2, Column 2, at the nearest potable water supply and in the nearest surface water supply in an Unrestricted Area.

E. Deleted F. Deleted G. Deleted E. Deleted I. Deleted J. Explosive Gas Mixture The hydrogen monitors are used to detect possible hydrogen buildups which could result in a possible hydrogen explosion. Automatic isolation of the off-gas flow would prevent the hydrogen explosion and possible damage to the augmented off-gas system. Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled.

K. Steam Jet Air Ejector (SJAE)

Restricting the gross radioactivity release rate of gases from the main condenser SJAE provides reasonable assurance that the total effective dose equivalent to an individual at the exclusion area boundary will not exceed the limits of 10CFR50.67 in the event this effluent is inadvertently discharged directly to the environment without treatment.

This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10CFR Part 50.

Amendment No. en, USE, GS-, 493, BEY 01 52, 223 175

k I VYNPS Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month/year) the change was implemented.

C. PRIMARY CONTAINMENT LEAKAGE RATE TESTING PROGRAM A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, entitled "Performance Based Containment Leak-Test Program," dated September 1995, as modified by the following:

  • The first Type A test after the April 1995 Type A test shall be performed no later than November 2005. (This is an exception to Section 9.2.3 of NEI 94-01, Rev. 0, "Industry Guideline for Implementing Performance-Based Option of 10CFR50, Appendix J.")

Option B; (2) Section 6.4.4 of ANSI/ANS 56.8-1994; and (3)

Section 10.2 of NEI 94-01, Rev. 0.

Option B; (2) Section 3.2 of ANSI/ANS 56.8-1994; and (3)

Sections 8.0 and 9.0 of NEI 94-01, Rev. 0.

The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 44 psig.

The maximum allowable primary containment leakage rate, La, at Pa, shall be 0.8% of primary containment air weight per day.

Leakage rate acceptance criteria are:

1. Primary containment leakage rate acceptance criterion c 1.0 La.
2. The as-left primary containment integrated leakage rate test (Type A test) acceptance criterion is < 0.75 La.
3. The combined local leakage rate test acceptance criterion for Type B and Type C tests (excluding the leakage contributions from the main steam pathways) is < 0.6 La, calculated on a maximum pathway basis, prior to entering a mode of operation where primary containment integrity is required.
4. The combined local leakage rate test acceptance criterion for Type B and Type C tests (excluding the leakage contributions from the main steam pathways) is < 0.6 La, calculated on a minimum pathway basis, at all times when primary containment integrity is required.

Amendment No. 1 45,412, 471, 2 223 26 5

I VYNPS

5. Airlock overall leakage rate acceptance criterion is 5 0.10 La when tested at 2 Pa.

The provision of SR for 4.0.2 for Surveillance Frequency does not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.

D. Radioactive Effluent Controls Program This program conforming to 10 CFR 50.36a provides for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by operating procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;
b. Limitations on the concentrations of radioactive material released in liquid effluents from the site to unrestricted areas, conforming to 10 times the concentration values in Appendix B, Table 2, Column 2, to 10 CFR 20.1001 - 20.2402;
c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents pursuant to 10 CFR 20.1302 and with the methodology and parameters in the ODCM;
d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from the unit to unrestricted areas, conforming to 10 CFR 50, Appendix I;
e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days;
f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2 percent of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I;
g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be limited to the following:
1. For noble gases: less than or equal to a dose rate of 500 mrems/yr to the total body and less than or equal to a dose rate of 3000 mrems/yr to the skin, and
2. For iodine-131, iodine-133, 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 mrems/yr to any organ; Amendment No. 141, 22-1, 223 26 6
Retrieved from "https://kanterella.com/w/index.php?title=ML050900399&oldid=2594578"