ML20140E182
| ML20140E182 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 01/27/1986 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20140E163 | List: |
| References | |
| TAC-61488, NUDOCS 8602030199 | |
| Download: ML20140E182 (34) | |
Text
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings 2.
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS, REACTOR CORE Applicability Applies to reactor thermal power, reactor power imbalance, reactor coolant system pressure, coolant temperature, and coolant flow during power operation of the plant.
Objective To maintain the integrity of the fuel cladding.
Specification 2.1.1 The combination of the reactor system pressure and coolant temperature 141>
shall not exceed the safety limit as defined by the Core Operating Limits Manual of Specification 6.9.1.1.
If the actual pressure /
temperature point is within the restricted region the safety limit is exceeded.
2.1.2 The combination of reactor thermal power and reactor power imbalance (power in the top half of the core minus the power in the bottom half of the core expressed as a percentage of the rated power) shall not exceed the safety limit as defined by the locus of points (solid ifne) for the 141><
specified flow set forth in the Core Operating Limits Manual.
If the actual-reactor-thermal-power / reactor-power-imbalance point is above the line for the specified flow, the safety limit is exceeded.
Bases The safety limits presented have been generated using BAW-2 and BWC CHF 141> correlations and the actual measured flow rate. The flow rate utilized is
< 104.9 percent of the design flow (369,600 gpm) based on four-pump operation.
To maintain the integrity of the fuel cladding and to prevent fission product release to the primary coolant system, it is necessary to prevent overheating i
of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling region of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is i
only sifghtly greater than the coolant temperature.
The upper boundary of the nucleate boiling region is termed " departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, tihich would result in high cladding temperatures and the possibility of cladding failure. Although DNB is not an observable parameter during reactor operation, the observable parameters of neutron power, reactor coolant flow, temperature, and pressure Proposed Amendment No. 141 2-1 8602030199 060127 ADOCKOSOOg2 DR
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings 141>< can be related to DNB through the use of the CHF correlation.
The BAW-2 and BWC correlations have been developed to predict DNB and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30 (BAW-2) or 1.18 (BWC).
A DNBR of 1.30 (BAW-2) or 1.18 (BWC) corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur; this is considered a conservative margin to DNB for all operating conditions.
The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits. The difference in these two pressures in nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip setpoints to correspond to the elevated location where the pressure is actually measured.
141> The curve presented in the Core Operating Limits Manual of reactor outlet
< temperature vs. core outlet pressure represents the conditions at which a DNBR equal to or greater than the correlation limit is predicted for the maximum possible thermal power (112 percent) when four reactor coolant pumps are operating (minimum reactor coolant flow is 104.9 percent of 369,000 gpm).
This curve is based on the combination of nuclear power peaking factors, with potential effects of fuel densification and rod bowing, which result in a more conservative DNBR than any other shape that exists during normal operation.
141> The combinations of reactor thermal power and reactor power imbalance
< presented in the Core Operating Limits Manual are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and rod bowing.
1.
The combinations of the radial peak, axial peak and position of the axial peak that yields a DNBR no less than the CHF correlation limit.
2.
The combination of radial and axial peak that causes central fuel melting at the hot spot.
The limit is 20.4 KW/ft.
Power peaking is not a directly observable quantity and therefore limits have been established on the bases of the reactor power imbalance produced by the power peaking.
141> The specified flow rates on the curve of thermal power level vs. reactor power
< imbalance in the Core Operating Limits Manual correspond to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.
141> For each of the curves in the Core Operating Limits Manual of reactor outlet temperature vs. core outlet pressure as determined by the RCS pump
< combinations, a pressure-temperature point above and to the lef t of the curve would result in a DNBR greater than the CHF correlation limit or a local quality at the point of minimum DNBR less than 22 percent for that particular 141> reactor coolant pump situation. A separate curve depicts the most restrictive 4 of all possible reactor coolant pump / maximum thermal power combinations.
Proposed Amendment No. 141 2-2
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings The maximum permitted thermal power for three-pump operation depicted in Figure 2.1-2 is 87.8 percent due to a power level trip produced by the flux-flow ratio 1.06 times 74.4 percent design flow = 78.86 percent power plus the absolute value of the maximum calibration and instrumentation error. The maximum thermal power for other coolant pump conditions is produced in a similar manner. The actual maximum power levels are calculated by the RPS and will be directly proportional to the actual flow during partial pump operation.
141><
Proposed Amendment No.141 2-3 I
1.
' RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings 1
141x Figure 2.1-1 This Figure has been deleted.
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings 141x Figure 2.1-2 This Figure has been deleted.
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings 141x Figure 2.1-3 This Figure has been deleted.
Proposed Amendment No. 141
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings B.
Pump Monitors The pump monitors prevent the minimum core DNBR from decreasing below the CHR correlation limit by tripping the reactor due to (a) the loss of two reactor coolant pumps in one reactor coolant loop, and (b) loss of one or two reactor coolant pumps during two-pump operation. The pump monitors also restrict the power level to 55 percent for one reactor coolant pump operation in each loop.
C.
Reactor coolant system pressure During a startup accident from low power or a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nuclear overpower trip set point. The trip setting limit shown in figure 2.3-1 for high reactor coolant system pressure (2300 psig) has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient (1) and minimize the challenges to the EH0V and code safeties.
The low pressure (1900 psig) and variable low pressure 141>
(12.96 Tout - 5834) trip set point presented in the Core Operating Limits Manual of Specification 6.9.1.1 has been established to maintain the DNB ratio greater than or equal to the CHF correlation limit fo-those design accidents that result in a pressure reduction.
(2,3)
Due to the calibration and instrumentation errors the safety analysis used a variable low reactor coolant system pressure trip value of
- 5884).
(12.96 Tout D.
Coolant outlet temperature The high reactor coolant outlet temperature trip setting limit (618 F) shown in figure 2.3-1 has been established to prevent excessive core coolant temperatures in the operating range. Due to calibration and instrumentation errors, the safety analysis used a trip set point of 620 F.
E.
Reactor Building pressure The high Reactor Building pressure trip setting limit (4 psig) provides positive assurance that a reactor trip will occur in the unlikely event of a steam line failure in the Reactor Building or a loss of coolant accident, even in the absence of a low reactor coolant system pressure trip.
F.
Shutdown bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system. The reactor protection system segments which can be bypassed are shown in Proposed Amendment No. 141 2-7
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Safety Limits and Limiting Safety System Settings l
141H Figure 2.3-2 This Figure has been deleted.
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.1.3 MINIMUM CONDITIONS FOR CRITICALITY Specifications 3.1.3.1 The reactor coolant temperature shall be above 525 F except for portions of low power physics testing when the requirements of Specification 3.1.8 shall apply.
3.1.3.2 Reactor coolant temperature shall be above DTT + 10 F.
3.1.3.3 When the reactor coolant temperature is below the minimum temperature specified in 3.1.3.1 above, except for portions of low power physics testing when the requirements of Specification 3.1.8 shall apply, the reactor shall be subcritical by an amount equal to or greater than the calculated reactivity insertion due to depressurization.
3.1.3.4 The reactor shall be maintained subcritical by at least 1 percent ak/k until a steam bubble is formed and an indicated water level between 10 and 316 inches is established in the pressurizer.
141>
3.1.3.5 Except for physics tests and as limited by the Core Operating Limits Manual of Specification 6.9.1.1, safety rod groups shall be fully withdrawn prior to any other reduction in shutdown margin by deboration or regulating rod withdrawal during the approach to critical ity.
Following safety rod withdrawal, the regulating rods shall be positioned within their position limits as defined by the 141><
Core Operating Limits Manual.
Bases At the beginning of life of the initial fuel cycle, the moderator temperature coefficient is expected to be slightly positive at operating temperatures with the operating configuration of control rods.
(1) Calculations show that above 525 F the positive moderator coefficient is acceptable.
Since the moderator temperature coefficient at lower temperatures will be less negative or more positive than at operating temperature, (2) startup and operation of the reactor when reactor coolant temperature is less than 525 F is prohibited except where necessary for low power physics tests.
The potential reactivity insertion due to the moderator pressure coefficient (2) that could result from depressurizing the coolant from 2185 psia to saturation pressure of 885 psia is approximately 0.1 percent ak/k.
During physics tests, special operating precautions will be taken.
In addition, the strong negative Doppler coefficient (1) and the small integrated ak/k would limit the magnitude of a power excursion resulting from a reduction of moderator density.
Proposed Amendment No. 141 3-6
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.2 HIGH PRESSURE INJECTION AND CHEMICAL ADDITION SYSTEMS Applicability Applies to the operational status of high pressure injection and chemical addition systems.
Objective To provide for adequate boration under all operating conditions to assure ability to bring the reactor to a cold shutdown condition.
Specification The reactor shall not remain critical unless the following conditions are met:
3.2.1 Two pumps capable of supplying high pressure injection are operable (also see Specification 3.3.2).
3.2.2 The borated water storage tank and its flow path to the reactor for high pressure injection are operable.
3.2.3 A source of concentrated boric acid solution in addition to the borated water storage tank is available and operable. This requirement is fulfilled by the concentrated boric acid storage tank.
This tank shall contain at least the equivalent of 10,000 gallons of 7,100 ppm boron.
System piping and valves necessary to establish a flow path for high pressure injection shall also be operable and shall,have at least the same temperature as the boric acid storage tank. One associated boric acid pump is operable. The concentrated boric acid storage tank water shall not be less than 70F, and at least one channel of heat tracing shall be operable for this tank's associated piping.
The concentrated boric acid storage tank boron concentration shall not exceed 8,500 ppa boron.
Bases The makeup and purification system and chemical addition s{ stems provide control of the reactor coolant system boron concentration.
This is normally accomplished by using either the makeup pump or one of the two high pressure injection purrps in series with a boric acid pump associated with the concentrated boric acid storage tank.
The alternate method of boration will be the use of the makeup or high pressure injection pumps taking suction directly from the borated water storage tank.2 The quantity of boric acid in storage from either of the two above-mentioned sources is sufficient to borate the reactor coolant system to a 1 percent subcritical margin in the cold condition (70F) at the worst time in core life 141> with a stuck contral rod assembly.
The minimum required gallonage, based on a boron concentration of 7100 ppm, is given in the Core Operating Limits Manual of Specification 6.9.1.1.
This requirement is satisfied by requiring a
< borated water volume of 7100 ppm in the concentrated boric acid Proposed Amendment No. 141 3-17
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 141> storage tank during critical operations in excess of the minimum requirement stated in the Core Operating Limits Manual.
The minimum volume for the borated water storage tank (390,000 gallons of 1800 ppm boron), as specified in section 3.3, is based on refueling volume requirements and easily satisfies the cold shutdown requirement. The specification assures that the two supplies are available whenever the reactor is critical so that a single failure will not prevent boration to a cold condition. The minimum volumes of boric acid solution given include the boron necessary to account for xenon decay.
The primary method of adding boron to the primary system is to pump the concentrated boric acid solution (7100 ppm boron, minimum) into the makeup tank using the 50 gpm boric acid pumps. Using only one of the two boric acid Jumps, the required volume of boric acid can be injected in less than 3.5 1ours.
The alternate method of addition is to inject boric acid from the borated water storage tank using the high pressure injection pumps.
Concentration of boron in the concentrated boric acid storage tank may be higher than the concentration which would crystallize at ambient conditions.
For this reason and to ensure that a flow of boric acid is available when needed, this tank and its associated piping will be kept above 70F (30F above the crystallization temperature for the concentration present). Once in the i
high pressure injection system, the concentrate is sufficiently well mixed and diluted so that normal system temperatures ensure boric acid solubility. The value of 70F is signigicantly above the crystallization temperature for a solution containing 12,200 ppm boron.
REFERENCES (1)
FSAR subsections 9.2 and 9.3.
(2)
FSAR Figure 6.2-1.
(3)
4 Proposed Amendment No. 141 t
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.2 Control Rod Group and Power Distribution Limits i
Applicability This specification applies to power distribution and operation of control rods during power operation.
Objective To assure an acceptable core power distribution during power operation, to set a limit on potential reactivity insertion from a hypothetical control rod ejection, and to assure core subcriticality after a reactor trip.
Specification 3.5.2.1 The available shutdown margin shall be not less than 1% ak/k with the highest worth control rod fully withdrawn.
If the shutdown margin is less than 1% ak/k then, within one hour, initiate and continue 141>
boration until the required shutdown margin as indicated in the Core Operating Limits Manual of Specification 6.9.1.1 is established.
3.5.2.2 Operation with inoperable rods:
A.
Operation with more than one inoperable rod as defined in Specification 4.7.1 and 4.7.2.3 in the safety or regulating rod banks shall not be permitted.
B.
If a control rod in the regulating and/or safety rod banks is declared inoperable in the withdrawn position as defined in Specification paragraph 4.7.1.1 and 4.7.1.3, an evaluation shall be initiated immediately to verify the existence of 1%
ak/k hot shutdown margin. Boration may be initiated to increase the available rod worth eitner to compensate for the worth of the inoperable rod or until the regulating banks are l
fully withdrawn, whichever occurs first.
C.
If within one hour of determination of an inoperable rod as defined in Specification 4.7.1, it is not determined that a 1%
ak/k hot shutdown margin exists combining the worth of the inoperable rod with each of the other rods, the reactor shall be brought to the hot standby condition until this margin is established.
D.
Following the determination of an inoperable rod as defined in Specification 4.7.1, all rods shall be exercised by a movement i
until indication is noted but not exceeding 2 inches within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised weekly until the rod problem is solved.
E.
If a control rod in the regulating or safety rod groups is declared inoperable per 4.7.1.2, power shall be reduced to 60%
of the thermal power allowable for the reactor coolant pump combination.
Proposed Amendment No. 141 3-31
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation F.
If a control rod in the regulating or axial power shaping groups is declarad inoperable per Specification 4.7.1.2, operation above 60% of rated power may continue provided the rods in the group are positioned such that the rod that was declared inoperabic is maintained within allowable group average position limits of Specification 4.7.1.2 and the 141><
withdrawal limits the Core Operating Limits Manual.
3.5.2.3 The worth of a single inserted control rod shall not exceed 0.65 percent ak/k at rated power or 1.0 percent ak/k at hot zero power except for physics testing when the requirement of Specification 3.1.8 shall apply.
3.5.2.4 Quadrant Power Tilt A.
With the Quadrant Power Tilt determined to exceed 4.92% but less than or equal to 11.07% except for physics test.
1.
Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
a) Either reduce the quadrant power tilt to <4.92%, or b) Reduce thermal power so as not to exceed thermal power, including power level cutoff, allowable for the reactor coolant pump combination, less at least 2% for each 1%,
or fraction thereof, of quadrant power tilt in excess of 4.92%. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, take action to reduce the high flux trip and flux-a flux-flow trip setpoints at least 2% for each 1%, or fraction thereof, of quadrant power tilt in excess of 4.92%.
2.
Verify that the Quadrant Power Tilt is <4.92% within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding that limit or reduce Thermal Power to less than 60% of Thermal Power allowable for the reactor coolant pump combination within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the High Flux Trip Setpoint to <65.5% of Thermal Power allowable for the reactor coolant pump combination within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
3.
Identify and correct the cause of the out of limit condition prior to increasing Thermal Power; subsequent Power Operation above 60% of Thermal Power allowable for the reactor coolant pump combination may proceed provided that the Quadrant Power Tilt is verified <4.92%
at least once per hour for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or until verifTed acceptable at 95% or greater Rated Thermal Power.
Proposed Amendment No.141 3-32
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RANCHO SECO UNIT 1 l
TECHNICAL SPECIFICATIONS Limiting Conditions for Operation E.
Whenever the reactor is brought to hot shutdown pursuant to 3.5.2.4.0 above, subsequent reactor operation is permitted for the purpose of measurement, testing and corrective action j
provided the thermal power and pcwer range high flux setpoint allowable for the Reactor Coolant Pump combination are restricted by a reduction of 2% of maximum allowable power for 1
6 each 1% tilt, or fraction thereof, for the maximum tilt observed prior to shutdown.
F.
The Quadrant Power Tilt shall be determined to be within the limits at least once every shift during operation above 15% of i
Rated Thermal Power except when the Quadrant Power Tilt alarm is inoperable, then the Quadrant Power Tilt shall be calculated and evaluated at least once every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
3.5.2.5 Control Rod Positions A.
Technical Specifiction 3.1.3.5 (safety rod withdrawal) does not i;
prohibit the exercising of individual safety rods as required by Table 4.1-2 or apply to inoperable safety rod limits in Technical 4
Specification 3.5.2.2.
B.
Operating rod group overlap shall be 25% *5% between three sequential groups, except for physics tests.
C.
Position limits are specified for regulating control rods.
Except for physics tests or exercising control rods, the regulating control rod 141>
insertion / withdrawal limits are specified in the Core Operating Limits l
Manual of Specification 6.9.1.1.
If any of these control rod position limits are exceeded, such that control rod positions are in the restricted region, an acceptable control rod position shall be obtained within two hours.
If control rod positions exceed the l
shutdown margin limit, such that control rods are in the region defined as operation not allowed then, within one hour, initiate and continue boration until the required shutdown margin is achieved.
D.
Except for physics test, power shall not be increased above the power level cutoff of 92% of the maximum allowable power level unless one of the following conditions is satisfied:
1.
Xenon reactivity is within 10% of the equilibrium value for operation at the maximum allowable power level and i -
asymptotically approaching stability.
l 2.
Except for Xenon-free startup, when 3.5.2.5D(1) applies, the reactor has operated within a range of 87 to 92% of the maxin:um allowable power for a period exceeding 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in the soluble poison control mode.
Proposed Amendment No.141 3-33
RANCHO SECO UNIT 1 4
TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.5.2.6 Reactor power imbalance shall be monitored on a frequency not to exceed two hours during power operation above 40 percent rated power.
Except for physics test, imbalance shall be maintained within Specification 6.9.1.1.y the Core Operating Limits Manual ofIf the imbalance is not the envelope defined b 141>
defined in the Core Operating Limits Manual, corrective measures shall be taken to achieve an acceptable imbalance.
If an acceptable imbalance is not achieved within two hours, reactor power shall be reduced until imbalance limits are met.
3.5.2.7 The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent or his designated representative.
Bases 141>< The power-imbalance envelope defined in the Core Operating Limits Manual are based on LOCA analyses which have defined the maximum linear heat rate such that the maximum clad temperature will not exceed the Final Acceptance 141>< Criteria.
Corrective measures will be taken should the indicated quadrant tilt, rod position, or inbalance be outside their specified boundry.
Operation in a situation that would cause the Final Acceptance Criteria to be approached should a LOC occur is highly improbable because all of the power distribution parameters (quadrant tilt, rod position, and imbalance) must be at their limits while simultaneously all other engineering and uncertainty factors are also at their limits.**
a.
Nuclear uncertainty factors b.
Thermal calibration c.
Hot rod manufacturing tolerance factors d.
Fuel densification effects The conservative application of the above peaking augmentation factors compensates for the potential peaking penalty due to Fuel rod bow.
141>< The 25%,i 5%
overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke.
Control rods are arranged in groups or banks defined as follows:
Group Function 1
Safety 2
Safety 3
Safety 4
Safety 5
Regulating 6
Regulating 7
Regulating 8
APSR (axial power shaping group)
Actual operating limits depend on whether or not incore or excore detectors are used and their respectivce instrument calibration errors.
The method used to define the operating limits is defined in plant operating procedures.
Proposed Amendment No. 141 3-33a
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The rod position limits are based on the most limiting of the following three criteria: ECCS power peaking, shutdown margin, and potential ejected rod worth. Therefore, compliance with the ECCS power peaking criterion is ensured by the rod position limits.
The minimum available rod worth, consistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod that is withdrawn remains in the full out position.Il? The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than 0.65 ak/k analysis of hypothetical rod ejection accident.t2ye safe by the safety at rated power. These values have been shown t9 A maximum single inserted control rod worth of 1.0 ak/k is allowed by the rod position limits at hot zero power. A single inserted control rod worth of 1.0 ak/k at beginning-of-life, hot zero power would result in a lower transient peak thermal power and, therefore, less severe environmental consequences than an 0.65 ak/k ejected rod worth at rated power.
Control rod groups are withdrawn in sequence beginning with Group 1.
Groups 5, 6 and 7 are overlapped 25 percent. The normal position at power is for Group 7 to be partially inserted.
The Quadrant Power Tilt limits set forth in Specification 3.5.2.4 have been established to prevent the linear heat rate peaking increase associated with a positive quadrant power tilt during normal power operation from exceeding 7.36.
The limits in Specification 3.5.2.4 are measurement system independent. The actual operating limits, with the appropriate allowance for observability and instrumentation errors, for each measurement system are defined in the station operating procedures.
The Quadrant Tilt and axial imbalance monitoring in Specifications 3.5.2.4F and 3.5.2.6, respectively, normally will be performed in the process computer.
The two-hour frequency for monitoring these qualities will provide adequate surveillance when the computer is out of service.
Allowance is provided for withdrawal limits and reactor power imbalance limits to be exceeded for a period of two hours without specification violation.
Acceptable rod positions and imbalance must be achieved within the two-hour time period or appropriate action such as a reduction of power taken.
141>< Operating restrictions are included in the Core Operating Limits Manual to prevent excessive power peaking by transient xenon. The xenon reactivity must either be beyond the "undershoot" region and asymptotically approaching its equilibrium value at rated power or the reactor must be operated in the range 141>< indicated in the Core Operating Limits Manual for a period exceeding two hours in the soluble poison control mode so that the transient peak is burned out at a lower power level.
REFERENCES (1) FSAR, Section 3.2.2.1.2 (2) FSAR, Section 14.2.2.4 Proposed Amendment No. 141 3-33b
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 141x Figure 3.5.2-1 This Figure has been deleted.
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Proposed Amendment No. 141
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 141x Figure 3.5.2-12 This Figure has been deleted.
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Proposed Amendment No. 141 4
-r RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.8 PROCEDURES (Continued) b.
The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.
c.
The change is documented, reviewed by the PRC and approved by the Plant Superintendent within seven (7) days of implementation.
6.9 REPORTING REQUIREMENTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted.
141>
6.9.1.1 Core Operating Limits Manual A manual providing the following core operational limits shall be provided to the Regional Administrator of the Regional Office of the NRC with a copy to the Director, NRR, Attention: Chief, Core Performance Branch, USNRC at least 60 days prior to each cycle initial criticality unless otherwise approved by the Commission by letter:
1.
Regulating Control Rod Insertion / Withdrawal Limits 2.
Axial Power Shaping Control Rod Insertion / Withdrawal Limits 3.
Power Level Cutoff 4.
Power / Power Imbalance Operational Limits 5.
RPS Maximum Allowable Setpoints In addition, in the event that the limits should change requiring a new submittal or amended submittal of the Core Operational Limits manual, it shall be submitted at least 60 days prior to the date the limit would become effective unless otherwise approved by the Commission by letter. Any information needed to support the Core Operational Limits Report will be requested from the NRC and need not be included in the manual.
Startup Report 141>< 6.9.1.2 A summary report of plant startup and power escalation testing shall be submitted following (1) Receipt of an operating license; (2) amendment to the license involving a planned increase in power level; (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier; and (4) modifications that may have significantly altered the nuclear, thermal or hydraulic performance of the plant. The report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other comitments shall be included in this report.
Proposed Amendment No.141 6-12
RANCHO SEC0 UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 141>< 6.9.1.3 Startup reports shall be submitted within (1) Ninety (90) days following completion of the startup test program; (2) Ninety (90) days following resumption or commencement of commercial power operation; or (3) Nine (9) months following initial criticality, whichever is earliest.
If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program and resumption or commencement of connerical power operation), supplementary reports shall be submitted at least every three (3) months until all three events have been completed.
6.9.2 Environmental Reports 6.9.2.1 Annual Radiological Reports Annual reports covering the activities of the unit, as described below, for the previous calendar year shall be submitted prior to Marth 1 of each year following initial criticality.
Reports required on an annual basis shall include:
1 A tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man rem exposure, according to work and job functions,*e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements. Small exposures, totaling less than 20 percent of the individual total dose, need not be accounted for.
In the aggregate, at least 80 percent of the total whole body dose received from external sources shall be assigned to specific major work functions.
6.9.2.2 Annual Radiological Environmental Operating Report 6.9.2.2.1 Routine radiological environmental operating reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year.
The initial report shall be submitted prior to May 1 of the year following initial criticality.
- This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.
Proposed Amendment No. 141 6-12a
RANCHO SEC0 UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.9.2.2.2 The annual radiological environmental operating reports shall include summaries, interpretations, and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate),
and previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the envi ronment. The reports shall also include the results of the land use censuses.
If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.
The annual radiological environmental operating reports shall include summarized and tabulated results in the format of Table 6.9-1, of all radiological environmental samples taken during the report period.
In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.
The missing data shall be submitted as soon as possible in a supplementary report.
The reports shall also include the following:
a summary description of the radiological environmental monitoring program; including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equipment used; a map of all sampling locations keyed to a table giving distances and directions from one reactor; the result of land use censuses, and the results of licensee participation in the Interlab Comparison Program. The annual report shall also include information related to Specification 4.29.
6.9.2.3 Semiannual Radioactive Effluent Release Report Routine radioactive effluent release reports covering the operation of the unit during the previous six months of operation shall be 4
submitted within 60 days after January 1 and July 1 of each year.
The period of the first report shall begin with the date of initial criticality.
6.9.2.3.1 The radioactive effluent release reports shall include a summary of the quantities of radioactive liquid and gaseous effluents t
and solid waste released from the unit as outlined in Regulatory l
Guide ic?l, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid j
and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," with data summarized on a quarterly basis, following the format of Appendix B thereof.
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Proposed Amendment No. 141 6-12b i
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls 6.9.2.3.1 Continued)
The radioactive effluent release reports shall include the release of gaseous effluents during each quarter, as outlined in Regulatory Guide 1.21, with the data summarized on a quarterly basis, following the format of Appendix B thereof.
A summary of meteorological conditions suring the release of gaseous effluents will be retained on-site for two years.
In addition, j
any changes to the Offsite Dose Calculation Manual will be submitted with the Semiannual Radioactive Effluent Release Report.
The radioactive effluent release reports shall include an assessment of the radiation doses from radioactive effluents to individuals due i
to their activities inside the site boundary during the report i
period. All assumptions used in making these assessments (e.g.,
i.
specific activity, exposure time, and location) shall be included in these reports.
The radioactive effluent release reports shall include the following information for all unplanned releases to unrestricted areas of j
radioactive materials in gaseous and liquid effluents:
a.
A description of the event and equipment involved.
Cause(s) for the unplanned release.
b.
1 c.
Actions taken to prevent recurrence.
d.
Consequences of the unplanned release.
The radioactive effluent release reports shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the unit during each calendar quarter, as outline in Regulatory Guide 1.21.
The releases of effluents shall be used for determining the gaseous pathway doses. The assessment of radiation i
doses shall be performed in accordance with the Offsite Dose Calculation Manual (0DCM).
i The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGkAM (PCP) or (00CM) made during the reporting period, as provided in Specifications 6.14 and 6.15.
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MONTHLY REPORT 6.9.3 Routine reports of operating statistics, including narrative summary of operating and shutdown experience, of lifts of the Primary System Safety Valves or EMOVs, of major safety related maintenance, and tabulations of facility changes, tests or experiments required pursuant to 10 CFR 50.59(b), shall be submitted on a monthly basis to the Office of Management Information and Program Control, U. S.
Nuclear Regulatory Commission, Washington, D. C. 20555, with a copy to the Regional Office, postmarked no later than the 15th day of each month following the calendar month covered by the report.
j Proposed Amendment No. 141 6-12c
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Administrative Controls LICENSEE EVENT REPORT 6.9.4 The LICENSEE EVENT REPORTS of Specification 6.9.4.1 below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC as Licensee Event Reports.
Supplemental reports may be required to fully describe final resolution of occurrence.
In case of corrected or supplemental reports, a License Event Report shall be completed and reference shall be made to the original report date, pursuant to the requirements of 10 CFR 50.73.
6.9.4.1 The types of events listed below shall be the subject of written reports to the Director of the Regional Office within thirty (30) days of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form, pursuant to 10 CFR 50.73 and the guidance of NUREG-1022.
a.
(i)
The completion of any nuclear plant shutdown required by the plant's Technical Specifications; or (ii) Any operation or condition prohibited by the plant's Technical Specifications; or (iii)
Any deviation from the plant's Technical Specifications authorized pursuant to 10 CFR 50.54(x).
b.
Any event or condition that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded, or that resulted in the nuclear power plant being:
(i)
In an unanalyzed condition that significantly compromised plant safety; (ii) In a condition that was outside the design basis of the plant; or (iii)
In a condition not covered by the plant's operating and emergency procedures.
c.
Any natural phenomenon or other external condition that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant.
d.
Any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF), including the Reactor Protection System (RPS). However, actuation of an ESF, including the RPS, that resulted from and was part of the preplanned sequence during testing or reactor operation need not be reported.
Proposed Amendment No. 141 6-12d
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RANCHO SECO UNIT 1
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TECHNICAL SPECIFICATIONS Administrative Controls LICENSEE EVENT REPORT (Continued) e.
Any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to:
1.
Shut down the reactor and maintain it in a safe shutdown condition; 2.
Remove residual heat; 3.
Control the release of radioactive material; or 4.
Mitigate the consequences of an accident.
f.
Events covered in paragraph 6.9.4.1.e of this section may include one or more procedural errors, equipment failures, and/or discovery of design, analysis, fabrication, construction, and/or procedural inadequacies. However, individual component failures need not be reported pursuant to this paragraph if redundant equipment in the same system was operable and available to perform the required safety function.
g.
Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to:
1.
Shut down the reactor and maintain it in a safe shutdown conditf or" 2.
Remove residual heat; 3.
Control the release of radioactive material; or 4.
Mitigate the consequences of an accident.
h.
1.
Any airborne radioactivity release that exceeded 2 times the applicable concentrations of the limits specified in Appendix B, Table II of 10 CFR 20 in unrestricted areas, when averaged over a time period of one hour.
2.
Any liquid effluent release that exceeded 2 times the limiting combined Maximum Permissible Concentration (MPC)
(see Note 1 of Appendix B to 10 CFR 20) at the point of entry into the receiving water (i.e., unrestricted area) for all radionuclides except tritium and dissolved noble gases, when averaged over a time period of one hour.
1.
Any event that posed an actual threat to the safety of the nuclear power plant or significantly hampered site personnel in the performance of duties necessary for the safe operation of the nuclear power plant including fires, toxic gas releases, or radioactive releases.
J.
Failure of the pressurizer EMOVs or Primary System Safety Valves.
Proposed Amendment No. 141 6-12e
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