ML18139B282
| ML18139B282 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 04/28/1981 |
| From: | Ferguson J Virginia Power (Virginia Electric & Power Co) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18139B283 | List: |
| References | |
| 232, NUDOCS 8105040191 | |
| Download: ML18139B282 (76) | |
Text
VIRGINIA ELECTJUC AND POWER COMPANY RicllMOND, VmGINIA 23261 JACK H.FERGUSON ExscUTIVE V:rcB PBEsroENT April 28, 1981 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation Attn:
Mr. Steven A. Varga, Chief Operating Reactors Branch No. 1 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C.
~0555 Gentlemen:
Serial No. 232 FR/RWC/WRM:ms Docket Nos. 50-280 50-281 License Nos. DPR~32 DPR-37 AMENDMENT TO OPERATING LICENSES DPR-32 AND DPR-37 SURRY POWER STATION UNITS NO. 1 AND NO. 2 PROPOSED TECHNICAL SPECIFICATIONS CHANGES Pursuant to 10 CFR 50. 59, the Virginia Electric and Power Company hereby requests amendment, in the form of changes to the Technical Specifications, to Operating Licenses DPR-32 and DPR-37 for the Surry Nuclear Power Station Units No. 1 and No. 2.
The proposed changes are enclosed.
The LOCA-ECCS analysis results provided in our letter dated May 31, 1979 from Mr. C. M. Stallings (VEPCO) to Mr. Harold R. Denton (NRC)
(Serial No. 388),
support the continued full rated power operation of both Surry Units 1 and 2 after replacement of their respective steam generators.
This LOCA-ECCS analysis was approved by your letter to Mr. J. H. Ferguson dated May 16, 1980.
The attached proposed amendment satisfies the commitment in our May 31, 1979 letter to provide the additional Technical Specifications changes required to support operation of Surry Unit 1 prior to the completion of the Surry Unit 1
Steam Generator Replacement Program.
Since both units will have replacement steam generators, the appropriate Technical Specifications have been recombined.
In addition, we have changed the total peaking factor value (F0 ) for both units from 2.19 to 2.18.
This formally implements the adminis-trative restriction we had imposed with our letter from Mr. B. R. Sylvia to Mr. Harold R. Denton dated July 28, 1980 (Serial No. 664) as a result of our further assessment of NUREG 0630. provides a safety evaluation which supports elimination of the requirement for frequent axial power distribution surveillance based on the maximum analytically predicted total peaking factor values for Cycle 6 of Surry Unit 1 which are less than the limit imposed by our July 28, 1980 letter referenced above. also supports a related modification of the Axial Flux Difference limits. provides the appropriate changes to the Technical Specifications.
f
- 1oo/
s
~~b/r t
..;;J /' I 6. c.
\\n / V',.
h f I oo.oO
- 1' *,.
\\".
VIRGINIA ELECTRIC AND POWER COMPANY TO 2
Numerous editorial changes, consisting primarily of grammatical corrections, are incorporated in this request in order to clarify the meaning and intent of both the Limiting Conditions for Operation and the Bases..Specifically, Specifications 3.12. C. 5 and 3.12.D have been reformated to clarify the requirements involved; however, the reformating does not alter the meaning of the requirements.
This request also deletes references to Specification 3.12.B.2.b.(2),
which no longer exists, in Specifications 3.12.B.1 and
- 6. 6. 2. a. (2).
An additional editorial change is the deletion of numerous "blank spaces" located between various requirements throughout Section 3.12.
These "blank spaces" resulted from the deletion of previous requirements by Staff issued L:i.cense Amendments.
This proposed amendment has been reviewed and approved by our Station Nuclear Safety and Operating Committee and our System Nuclear Safety and Operating Committee.
It has been determined that this request does not involve an unreviewed safety question as defined in 10 CFR 50.59.
We have evaluated this request in accordance with the criteria in 10 CFR 170.22.
Since this request involves a safety issue for Unit 1 which the staff should be able to determine does not represent a significant hazards consider-ation and involves a duplicate request for Unit 2, a Class II license amendment fee and a Class I license amendment fee is required for Unit 1 and Unit 2, respectively.
Accordingly, a voucher check in the amount $1,600.00 is enclosed in payment of the required fees.
Your review of the enclosed Technical Specifications changes is requested by May 15, 1981.
Should you have questions, we would be happy to discuss this with you at your convenience.
Enclosures
- 1.
Safety Evaluation
- 2.
Proposed Technical Specification Change
- 3.
Voucher Check No.. 19264 for $1,600.00 cc:
Mr. James P. O'Reilly, Director Office of Inspections and Enforcement Region II Very truly yours, J. H. Ferguson Executive Vice President Power
~--
COMMONWEALTH OF VIRGINIA)
)
CITY OF RICHMOND
)
The foregoing document was acknowledged before me, in and for the City and Commonwealth aforesaid, today by J.. H.
- Ferguson, who is Executive Vice President-Power, of the Virginia Electric and Power Company.
He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.
-tL Acknowledged before me this.:2 >>
day of 19 2).
My Commission expires: ---~_;:.__-_~_t __, 19 J'.S-.
Notary Public (SEAL)
ATTACHMENT 1 SAFETY EVALUATION (TOTAL PEAKING FACTOR) FOR SURRY UNIT NO. 1 The analytically predicted maximum values for the total peaking factor, Fq (Z), were determined using the "3-ca_se" analysis methodology documented in Reference 1.
These predicted total peaking factor values lie below the Technical Specifications limit of 2.19 X K(Z) which is described in Reference 2, and the limit of 2.18 which we administratively imposed by Reference 3.
Thus, no potential violations of the proposed Technical Speci-fications limit for Fq(Z) exist during Condition I plant operation in Surry 1, Cycle 6, and frequent axial power distribution surveillance is not necessary.
The axial flu..~ difference (AFD) limits applicable for Cycle 6 are the generic values developed by Westinghouse based upon the Constant Axial Offset Control (CAOC) analysis methodology (+/-5~1 band) which is d~scribed in WCAP-83854 and in Reference 1.
The predicted values for Fq(Z) wer~ computed using these generic. Westinghouse AFD limits in the specific CAOC analysis for*Surry 1, Cycle 6.
l, i
References
- 1.
Letter from Westinghouse (C. Eicheldinger) to NRC (J. F. Stolz) dated April 6, 1978, Serial No. NS-CE-1749.
- 2.
Letter from Vepco (C. M. Stallings) to NRC (H. R. Denton) dated May 31, 1979, Serial No. 388.
- 3.
Letter from Vepco (B. R. Sylvia) to NRC (H. R. Denton) dated July 28, 1980, Serial No. 664.
- 4.
T. Morita, et. al., "Power Distribution Control and Load Following Procedures, "WCAP-8385, Westinghouse Electric Corporation, September, 1974.
ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION CHANGE SURRY POWER STATION
7 0
e TS 3.12-1 3.12 CONTROL ROD ASSEMBLIES AND POWER DISTRIBUTION LIMITS Applicability Applies to the operation of the control rod assemblies and power distri-bution limits.
Objective To ensure core subcriticality after a reactor trip, a limit on potential reactivity insertions from hypothetical control rod assembly ejection, and an acceptable core power distribution during power operation.
Specification A.
Control Bank Insertion Limits
- 1.
Whenever the reactor is critical, except for physics tests and control rod assembly exercises, the shutdown control rods shall be fully withdrawn.
- 2.
Whenever the reactor is critical, except for phrsics tes~~ and control rod assembly exercises, the full length control rod banks shall be inserted no further than the appropriate lJ.111it determined by core,burnup shown on TS Figures 3.12-lA, 3.12-lB, 3.12-2, or 3.12-3 for three-loop operation and TS Figures 3.12-4A, 3.12-4B, 3.12-5 or 3.12-6 for two-loop operation.
- 3.
The limits shown on TS Figures 3.12-lA through 3.12-6 may be revised on the basis of physics calculations and physics data obtained during unit startup and subsequent operation, in accordance with the following:
- a.
The sequence of withdrawal of the controlling banks, when going from zero to 100% power, is A, B, C, D.
- b.
An overlap of control banks, consistent with physics cal-
TS 3.12-2 culations and physics data obtained during Unit Startup and subsequent operation, will be permitted.
- c.
The shutdown margin with allowance for a stuck control rod assembly shall be greater than or equal to 1.77% reactivity under all steady-state operation conditions, except for physics tests, from zero to full power, including effects of axial pow~r distribution.
The shutdown margin as used here is defined as the amount by which the reactor core would be subcritical at hot shutdown conditions (T
~547°F) if all control rod avg assemblies were tripped, assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xenon or boron.
- 4.
Whenever the reactor is subcritical, except for physics tests, the critical rod position, i.e., the rod position at which criticality would be achieved if the control rod assemblies were withdrawn in normal sequence with no other reactivity changes, shall not be lower than the insertion _limit for zero power.
- s.
Insertion limits do not apply during physics tests or during periodic exercise of individual rods.
However, the shutdown margin indicated above must be maintained except for the low power physics test to measure control rod worth and shutdown margin.
For this test the reactor may be critical with all but one full control rod, expected to have the highest worth, inserted.
I.
B.
TS 3.12-3 Power Distribution Limits
- 1.
At all times except during low power physics tests, the hot channel factors defined in the basis must meet the following limits:
FQ(Z) ~ 2.18/P x K(Z) for P > 0.5 FQ(Z) ~ 4.36 x K(Z) for P ~ 0.5 FN
~ 1.55 (1+0.2(1-P))
Ml where Pis the fraction of rated power at which the core is operating, K(Z) is the function given in TS Figure 3.12-8, and Z is the core height location of FQ.
- 2.
Prior to exceeding 75% power following each core loading and during each effective full power month of operation thereafter, power distri-bution maps using the movable detector system shall be made to confirm that the hot channel factor limits of this specification are satis-fied.
For the purpose of this confirmation:
- a.
The measurement of total peaking factor ~eas shall be increased by eight percent to account for manufacturing tolerances, measure-ment error and-the effects a£ rod bow.
The measurement of enthalpy*
rise hot channel factor FM! shall be increased by four percent to account for measurement error. If any measured hot channel factor exceeds its limit specified under Specification 3.12.B.1, the reactor power and high neutron flux trip setpoint shall be reduced until the limits under Specification 3.12.B.1 are met.
If the hot channel factors cannot be brought to within the limits of FQ(Z)
~ 2.18 x K(Z) and ~Ml ~ 1. 55 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the Overpower 6.T and Overtemperature 6.T trip setpoints shall be similarly reduced.
- 3.
- TS 3.12-4 The reference equilibrium indicated axial flux difference (called the target flux difference) at a given power level P is that 0
indicated axial flux difference with the core in equilibrium xenon conditions (small or no oscillation) and the control rods more.than 190 steps withdrawn.
The target flux difference at any other power level Pis equal to the target value at P multiplied by the ratio 0
P/P.
The target flux difference shall be measured at least once per 0
equivalent full power quarter.
The target flux difference must be
- updated during each effective full power month of operation either by actual measurements or by linear interpolation using the most recent value and the value predicted for the end of the cycle life.
- 4.
Except as modified by Specifications 3.12.B.4.a, b, c, or d below, the indicated.axial flux difference shall be maintained within a
!5% band about the target flux difference (defines the target band on axial flux difference).
- a.
At a power level greater than. 90 percent of rated powe*r, if
- the indicated axial flm:c difference deviates from its target band, within 15 minutes either restore the indicated axial flux difference to within the target band or reduce the reactor power to less than 90 percent of rated power.
- b.
At a power level no greater than 90 percent of rated power, (1)
The indicated axial flux difference may deviate from its target band for a maximum of one hour (cumulative) in any 24-hour period provided the flux difference is within the limits shown on TS Figure 3.12-10.
- TS 3.12'."5 One minute penalty is accumulated for each one minute of operation outside of the target band at power levels equal to or above 50% of rated power.
(2) If Specification 3.12.B.4.b(l) is violated, then the reactor power shall be reduced to less than 50% power within 30 minutes and the high neutron flux setpoint shall be reduced to no greater than 55% power within the next four hours.
(3)
A power increase to a level greater than 90 percent of rated power is contingent upon the indicated axial flux difference being within its target band.
(4)
Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to TS Table 4.1-1 provided the indicated axial flux difference is maintained within the limits of TS Figure 3.12-10.
A total of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> of operation may be accumulated with the axial flux difference outside of the tar~et band during-this tes~ing without penalty deviation.
- c.
At a power level no greater than 50 percent of rated power, (1)
The indicated axial flux difference may deviate from its target band.
(2)
A power increase to a level greater than 50 percent of rated power is contingent upon the indicated axial flux difference not being outside its target band for more than one hour accumulated penalty during the preceding 24-hour period.
One half minute penalty is accumulated for each one minute of operation outside of the target band at power levels between 15% and 50% of rated power.
- TS 3.12-6
- d.
The axial flux difference limits for Specifications 3.12.B.4.a, b, and c may be suspended during the performance of physics tests provided:
(1)
The power level is maintained at or below 85% of rated power, and (2)
The limits of Specification 3.12.B.1 are maintained.
The power level shall be determined to be less than or equal to 85% of rated power at least once per hour during physics tests. Verification that the limits of Specification 3.12.B.1 are being met shall be demonstrated through in-core flux mapping at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Alarms shall normally be used to indicate the deviations from the axial flux difference requirements in Specification 3.12.B.4.a and the flux difference time limits in Specifications 3.12.B.4.b and c.
If the alarms are out **of service temporarily, the axial flux difference shall be logged and conformance to the limits assessed every hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and half-hourly thereafter.
The indicated axial flux difference for each excore channel shall be monitored at least once per 7 days when the alarm is operable and at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the alarm to operable status.
- 5.
The allowable quadrant to average power tilt is 2.0%.
- 6.
If, except for physics and rod exercise testing, the quadrant to average power tilt exceeds 2%, then:
I
- 1.
TS 3.12-7
- a.
The hot channel factors shall be determined within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the power level adjusted to meet the requirement of Specifi-cation 3.12.B.1, or
- b.
If the hot channel factors are not determined within two hours, the power level and high neutron flux trip setpoint shall be reduced from rated power 2% for each percent of quadrant tilt.
- c.
If the quadrant to average power tilt exceeds ~10%, the power level and high neutron flux trip setpoint will be reduced from rated power 2% for each percent of q~adrant tilt.
- 7.
If, except for physics and rod exercise testing, after a further period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the power tilt in Specification 3.12.B.5 above is not corrected to less than 2%:
- a.
If design hot channel factors for rated power are not exceeded, an evaluation as to the cause of the discrepancy shall be made and reported as a reportable occurrence to the Nuclear Regulatory Commission.
- b.
If the design hot channel factors for rated power are exceeded and the power is greater than 10%, the Nuclear Regulatory Commission shall be notified and the Nuclear Overpower, Nuclear Overpower llT, and Overtemperature llT trips shall be reduced one percent for each percent the hot channel factor exceeds the rated power design values.*
- c.
If the hot channel factors are not determined the Nuclear Regulatory Commission shall be notified and the Overpower
- c.
- TS 3.12-8
~T and Overtemperature ~T trip settings shall be reduced by the equivalent of 2% power for every 1% quadrant to average power tilt.
- 1.
A control rod assembly shall be considered inoperable if the assembly cannot be moved by the drive mechanism or the assembly remains misaligned from its bank by more than 15 inches.
A full-length control rod shall be considered inoper_able if its rod drop time is greater than 1.8 seconds to dashpot entry.
- 2.
No more than one inoperable control rod assembly shall be per-mitted when the reactor is critical.
- 3.
If more than one control rod assembly in a given bank is out of service because of a single failure external to the individual rod drive mechanism, i.e. programming circuitry, the provisions of Specifications 3.12.C.l and 3.12.C.2 shall not apply and the reactor may remain critical for a period not to exceed two hours provided immediate attention is directed.toward making the necessary repairs.
In the event the affected assemblies cannot be returned to service within this specified period the reactor will be brought to hot shutdown conditions.
I
- 4.
The provisions*of Specifications 3.12.C.1 and 3.12.C.2 shall not apply I during physics tests in which the assemblies are intentionally misaligned.
- 5.
The insertion limits in TS Figure 3.12-2 apply:
- a.
If an inoperable full-length rod is located below the 200 step level and is capable of being tripped, or
e
- b.
If the full-length rod is located below the 30 step level, whether or not it is capable of being tripped.
TS 3.12-9
- 6.
If an inoperable full-length rod cannot be located or if the inoperable full-length rod is located above the 30 step level and cannot be tripped, then the insertion limits in TS Figure 3.12-3 apply.
- 7.
If a' full-length rod becomes inoperable and reactor operation is continued, the potential ejected rod worth and associated transient power distribution peaking factors shall be determined by analysis within 30 days.
The analysis shall include due allowance for non-uniform fuel depletion in the neighborhood of the inoperabl~ rod. If the analysis results in a more limiting hypothetical transient than the cases reported in the safety analysis, the unit power level shall be'"reduced to an analytically determined part power level which is consistent with the safety analysis.
D.
Core Quadrant Power Balance:
- 1.
If the reactor is operating above 75% of rated power with one excore nuclear channel out of service, the core quadrant power balance shall be determined:
- a.
Once per day, and
- b.
After a cha~ge in power level greater than 10% or more than 30 inches of control rod motion.
E.
e TS 3.12-10
- 2.
The core quadrant power balance shall be determined by one of the following methods:
- a.
Movable detectors (at least two per quadrant)
- b.
Core exit thermocouples (at least four per quadrant)
Inoperable Rod Position Indicator Channels
- 1.
If a rod position indicator channel is out of service, then:
- a.
For operation between 50% and 100% of,rated power, the position of the RCC shall be checked indirectly by core instrumentaton (excore detector and/or thermocouples and/or movable incore detectors) every shift or subsequent to motion of the non-indicating rod exceeding 24 steps, whichever occurs first.
- b.
During operation below 50% of rated power, no special moni-toring is required.
- 2.
Not more than one rod position indicator (RPI) channel per group nor two RPI channels per bank shall be.permitted to be inoperable at any time.
F.
Misaligned or Dropped Control Rod
- 1.
If the Rod Position Indicator Channel is functional and the associated full length control rod is more than 15 inches out of alignment with its bank and cannot be realigned, then unless the hot channel fac;:tors are shown to be within design limits as specified in Specification 3.12.B.1 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, power shall be reduced so as not to exceed 75% of permitted power.
Basis e
TS 3.12-11
- 2.
To increase power above 75% of rated power with a full-length control rod more than 15 inches out of alignment with its bank, an analysis shall first be made to determine the hot channel factors and the resulting allowable power level based on Section 3.12-B.
The reactivity control concept assumed for operation is that reactivity changes accompanying changes in reactor power are compensated by control rod assembly motion. Reactivity changes associated with xenon, samarium, fuel depletion, and large changes in reactor coolant temperature (operating temperature to cold shutdown) are compensated for by changes in the soluble boron concen-tration.
During power operation, the shutdown groups are fully withdrawn and control of power is by the control groups.
A reactor*trip occurring during power operation will place the reactor into the hot-shutdown: condition.
The control rod assembly insertion limits provide for achieving hot shutdown by reactor trip at any time, assuming the high~st worth control rod assembly remains fully withdrawn, with sufficient margins to meet the assumptions used in the accident analysis.
In addition, they provide a limit on the maximum inserted rod worth in the unlikely event of a hypothetical assembly ejection and provide for acceptable nuclear peaking factors.
The limit may be deter-mined on the basis of unit startup and operating data to provide a more realistic limit which will allow for more flexibility in unit operation and
TS 3.12-12 still assure compliance with the shutdown requirement.
The maximum shut-down margin requirement occurs at end of core life and is based on the value used in the analysis of the hypothetical steam break accident.
The rod insertion limits are based on end of core life conditions.
The shut-down margin for the entire cycle length is established at 1.77% reactivity.
All other accident analysis with the exception of the chemical and volume control system malfunction analysis are based on 1% reactivity shutdown margin.
Relative positions of control rod banks are determined by a specified control rod bank overlap.
This overl?p is based on the consideration of axial power shape control.
The specified control rod insertion limits have been revised to limit the potential ejected rod worth in order to account for the effects of fuel densification.
The various control rod assemblies (shutdown banks,- control banks -A, B, C, and D) are each to be moved as a bank; that is, with all assemblies -in the bank within one step (5/8 inch) of the bank position.
Position indication is provided by two methods:
a digital count of actuating pulses which shows the demand position of the banks, and a linear position indicator, Linear Variable Differential Transformer, which indicates the actual assembly position.
The position indication accuracy of the Linear Differential Transformer is approximately !5% of span
(~ 7.5 inches) under steady state conditions.
The relative accuracy of the linear position indicator is such that, with the most adverse errors, an alarm is actuated if any two assemblies within a bank deviate by more than 14 inches.
In the event that the linear poisition indicator is not
e TS 3.12-13 in service, the effects of m~lpositioned control rod assemblies ~re obser-able from nuclear and process information displayed in the Main Co~trol Room and by core thermocouples and in-core movable detectors.
Below 50% power, no special monitoring is required for malpositioned control rod assemblies with inoperable rod position indicators because, even with an unnoticed complete assembly misalignment (full length control rod assembly 12 feet out of*align-ment with its bank), operation at 50% steady state power does not result in exceeding core limits.
The specified control rod assembly drop time is consistent with safety analyses that have been performed.
An inoperable control rod assembly imposes additional demands on the operators.
The permissible number of inoperable control rod assemblies is limited to one in order to limit the magnitude of the operating burden, but such a failure would not prevent dropping of the operable control rod assemblies upon reactor trip.
Two criteria have been chosen as a design basis for fuel performance related to fission gas release, pellet temperature, and cladding mechanical properties.
First, the peak value of fuel centerline temperature must not exceed 4700°F.
Second, the minimum_DNBR in the core must not be less than 1.30 in normal operation or in short term transients.
TS 3.12-14 In addition to the above, the peak linear power density and the nuclear enthalp~
rise hot channel factor must not exceed their limiting values which result from the large break loss of coolant accident analysis based on the ECCS acceptance criteria limit of 2200°F on peak clad temperature.
This is required to meet the initial conditions assumed for the loss of coolant accident.
To aid in specifying the limits of power distribution, the following hot channel factors are defined:
FQ(Z), Height Dependent Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerance on fuel pellets and rods.
E FQ' Engineering Heat Flux Hot Channel Factor, is defined as the allowance on heat flux required for manufacturing tolerances.
The engineering factor allows for local variations in enrichment, pellet density and diameter, surface area.
of the fuel rod, and eccentricity of the gap between pellet and clad.
Combined j ~
statistically the net*effect is a factor of 1.03 to be applied to fuel rod surface heat flux.
~' Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power for both LOCA and non-LOCA considerations.
e TS 3.12-15 It should be noted that the enthalpy rise factors are based on intergrals and are used as such in the DNB and LOCA calculations.
Local heat fluxes are obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in radial (x-y) power shapes throughout the core.
Thus, the radial power ~hape at the point of maximum heat flux.is not necessarily directly related to the enthalpy rise factors.
The results of the loss of coolant accident analyses are conservative with respect to the ECCS acceptance criteria as specified in 10 CFR 50.46 using an upper bound envelope of 2.18 times the hot channel factor normalized operating envelope given by TS Figure 3.12-8.
When an FQ measurement is taken, measurement error, manufacturing tolerances, and the effects of rod bow must be allowed for.
Five percent is the appropriate allowance for measurement error for a full core map (~40 thimbles monitored) taken with the movable incore detector flux mapping system, three percent is the appropriate allowance for manufactur'ing tolerances; *and five per-cent is the appropriate allowance for rod bow.
These uncertainties are statistically combined and result in a net increase of 1.08 that is applied to the measured value of FQ.
In the specified limit of~ there is an eight percent allowance for uncer-tainties, which means that normal operation of the core is expected to result in~~~ 1.55 (1+0.2 (1-P))/1.08.
The logic behind the larger uncertainty in this case is that (a) normal perturbations in the radial power shape (e.g., rod misalignment) affect~~' in most cases without necessarily affecting FQ' (b) the operator has a direct influence on FQ through movement of rods and can limit it to the desired value; he has no direct control over~~' and (c) an error in the predictions for radial power shape, which may be detected during startup physics tests and which may influence FQ' can
/
TS 3.12-17 between the top and bottom halves of two-section excore neutron detectors.
The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and the bottom halves of the core.
The permitted relaxation in~ with decreasing power level allows radial power shape changes with rod insertion to the insertion limits. It has been determined that provided the above conditions 1 through 4 are observed, this hot channel factor limit is met.
A recent evaluation of DNB test data obtained from experiments of fuel rod bowing in thimble cells has identified that the reduction in DNBR due to rod bowing in thimble cells is more than completely accommodated by existing thermal margins in the core design.
Therefore, it is not nec-essary to continue to apply a rod bow penalty to ~.6ij.
The procedures for axial power distribution contro~ are 9e~igned ~o mini-mize the effects of xenon redistribut"ion on the axial power distribution during load-follow maneuvers.
Basically, control of flux difference is required to limit the difference betwe~n the current value of flux dif-ference (al) and a reference value which corresponds to the full power equilibrium value of axial offset (axial offset= ~I/fractional power).
The reference value of flux difference varies with power level and burnup, but expressed as axial offset it varies only with burnup.
The technical specifications on power distribution control given in Specification 3.12.B.4 together with the surveillance requirements given in Specification 3.12.B.2 assure that the Limiting Condition for Operation for the heat flux hot channel factor is met.
TS 3.12-16 be compensated for by tighter axial control.
Four percent is the appropriate allowance for'measurement uncertainty for~ obtained from a full core map
(~40 thimbles monitored) taken with the movable incore detector flux mapping system.
Measurement of the hot channel factors are required as part of startup physics tests, during each effective full power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to
\\
a level based on measured hot channel-factors.
The incore map taken following core loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns.
The periodic incore mapping provides additional assurance that the nuclear design bases remain inviolate and identify opera-tional anomalies which would, otherwise, affect these bases.
For normal operation, it has been determined that, provided certain condi-tions are observed, the enthalpy rise hot channel factor~~ limit will be met.
These conditions are.as follows:
- 1.
Control rods in a single bank move together with-no individual rod insertion differing by more than 15 in~hes from the bank demand position.
An indicated misalignment limit of 13 steps precludes a rod misalignment no greater than 15 inches with consideration of maximum instrumentation error.
- 2.
Control rod banks are sequenced with overlapping banks as shown in TS Figures 3.12-lA, 3.12-lB, and 3.12-2.
- 3.
The full length control bank insertion limits are not violated.
- 4.
Axial power distribution control procedures, which are given in terms of flux difference control and control bank insertion limits are observed.
Flux difference refers to the difference
TS 3.12-18 The target (or reference) value of flux difference is determined as follows.
At any time that equilibrium xenon conditions have been estab-lished,"the indicated flux difference is noted with the full length rod control bank more than 190 steps withdrawn (i.e., normal full power opera-ting position appropriate for the time in life, usually withdrawn farther as burnup proceeds).
This value, divided by the fraction of full power at which the core was operating, is the full power value of the target flux difference~
Values for all other core power levels are obtained by multiplying the full power value by the fractional power.
Since the indi-cated equilibrium value was noted, no allowances for excore detector error are necessary and indicated deviation of ~5% ~I are permitted from the indicated reference value.
During periods where extensive load following is required, it may be impractical to establish the required core conditions for measuring the target flux difference every month.
For this reason, the specification provides two methods for updating the target flux difference.
Strict control of the flux difference (and rod position) is not as neces-*
sary during part power operation.
This is because xenon distribution control at part power is not as significant as the *control at full power and allowance has been made in predicting the heat flux peaking factors for less strict control at part power.
Strict control of the flux difference is not always possible during certain physics tests or during excore detector calibrations. Therefore, the specifications on power distribution control are less restrictive during physics tests and excore detector calibrations; this is acceptable due to the low probability of a significant accident occurring during these operations.
TS 3.12-19 In some instances of rapid unit power reduction automatic rod motion will cause the flux difference to deviate from the target band when the reduced power level is reached.
This does not nec*essarily affect the xenon dis-tribution sufficently to change the envelope of peaking factors which can be reached on a subsequent.return to full power within the target band; however, to simplify the specification, a limitation of one hour in any period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is placed on operation outside the band.
This ensures that the resulting xenon distributions are not significantly different from those resulting from operation within the target band.
The instantaneous consequences of being outside the band, provided rod insertion limits are observed, is not worse than a 10 percent increment
\\
in peaking factor for the allowable flux difference at 90% power, in.the range!. 13.8 percent (!10.8 percent indicated) where for every 2 percent below rated power, the permissible flux difference boundary is extended by 1 percent.
As discussed above, the essence of the procedure is to maintain the xenon distribution in the core as close to *the equilibrium full power condition as possible.
This is*accomplished, py using the boron system to position the full length control rods to produce the required indicated flux dif-ference.
A 2% quadrant tilt allows that a 5% tilt ~ight actually be present in the core because of insensitivity of the excore detectors for disturbances near the core center such as misaligned inner control rod and an error allowance.
No increase i:q FQ occurs with tilts up to 5% because* misaligned control rods* producing such tilts do not extend to the unrodded plane, wher~ the maximumFQ occurs.
{
a.a 0.2 Q
r:i t
~
co
- z:
0.4 1-1
- z:
Q 1-1 e--
I:;)
~
- .. -a 0.6 1-1 e,..
1-1 en Q
i:i.
~
- a a.a
=
l.O
- BANK C BANK D a.a 0.2 0.4 0.6 FRAC'l:ION OF RA'lEJ POWER e
TS li'IGilll 3.12-1.A 6-30-78 o.a 1.0 FIGO!U: 3.12-lA.CON'IROL BANK INSE1U:ION LIMITS FOR 3-LOOP NOBMAL OPE:RA.!ION-UNI'J: l Unit No. l Amendment rto. 42 Unit No. 2 Amendment No. 41 54
o.o 0.2
-a ta
...:-.-J, lil:J en
- z:
M
~
0.. 4 M t
= -
- z:
Q 0.*6 M
E-'4 M en 0
i:i..
~
=
a.a -
-/
/. t~*
1.0 o.o
- r- -
":A
~ -
.--- Bank C 7 -
- - Bank D
'/
- /
~
e-,-
0.2 0.4 0.6
!'RAC'?ION OF RAn:D POWU e
o.a TS FIGURE 3.12-l.B 10-8-77 45 l.O FIGURE 3.12-l.B CON'IIlOL BANK INSERTION Ln!ITS FOR.
Unit l Amendment No. 33 Unit 2 Amendment No. 32 NORMAL 3 LOOP OPERATION - UNLT Z
Q L,J,_
~
w U'l z -
z 0
I.
Q i
<C
°'
I l,;,.
0 I-
.* U'l 0
Q.
~
z
<C Cl a.a 0.2 0.4 0.6 0.8 e
BANK C
. (0.43)
CHANGE NO. 19 (0.47)
TS FIGURE 3.12-2 12~27-74 (0.09) 1
- a ~---.....J--------i--------"----....... -----.i a.a 0.2 0.4 0.6 1
- a FRACTION OF RATED POWER FIGURE 3.12-2 CONTROL BANK INSERTION Ll~dTS FOR 3 LOOP OPERATION WITH ONE BOTTO~ED ROD*
CHANGE NO. 19
(*
0 L&J 0.0 0.2 Ii:* 0. 4 L&J en
~
z 0....
~ 0.6 fE.
0.8 V
/
II' 1.0 0.0
/
/
e*
CHANGE NO. 9 f!GURE 3.12-3 CONTROL 8:\\:-:K DlSERnn:-: 1.1:!ITS FOR 3 LOOP OPERATIO::
WITH ONE r:;nPERABLE ROD V
/
I e
V
/ ~ BANK~
.v
/
I
/
I V
I I I VI I
I I I 1/1 I I I I
I I I
/
I I
I I
I
/
I I
i i I
I/
I I I
!./
I BANKO i I
I//.
i T
I I
I I I 1/1 I I I I*
1/l
.* j -
I I
/
I I
.I I
I I
I !/f
. I I
./ I I
I I
/
I I
I
. I I
I I
0.2 0.4 0.6 0.8 FRACTION FULL PO':.'ER ClL\\.~GE NO. i I/
- j.
I I I
I i
i I
TS FIGURE 3.12-3 8-~-73 I
lz
. w-18 --
l~j
?-
Io
~
I I~-
- Io I
I I
1.0
Q
!.I.I...
a::
w II) z z
0 I-(J a::
i... -z 0
. I-Ul 0 a..
~
z <
cc CHANGE NO
- 19 a.a (0.25) 0.2
( 0. 25) 0.4 0.6 a.a (0.80)
(0.19)
D TS rlGURE 3.12-4A 12-27-74
"'q:;
1.0..__~~~~--~~~~--~~~~----
a.a 0.2 0.4 0.6 FRACTION Or RATED POWER rlGU~E 3.12-4A CONTROL BANK INSERTION LIMITS FOR 2-LOOP NORMAL OPERATION UNIT NQ. 1 CHANGE No.* 19
(
1w t-
~
a.a 0.2 1w O. 4 in z
z 0
t-o
/
~ o. 6.
i..
/
0.8 1.0 a.a
,./
/
..., ~BA N~ C
~ V
/
V
/'
l/
8, ~NI D V
'-V
,!/"
V
~
/"
V 0.2 0.4 0.6 FRACTION OF' RATED POWER TS F'IGURE 3.12-48 6-10-75
I
\\_
C LaJ.
0.0 0.2 a: 0.4 LaJ en z z
0 t;
<- 0.6 fE 0.8 1.0 e
CHANGE ~O. 9 CONTROL BANK IXSERT!ON LIMITS FOR 2 LOOP OPERATION
~I!H or=E BO!TQ;',rEJl ROD I A" I
/f I I
/i I j
i I
I i
i i
I 8.:\\Ni<C I
- /
I i
/i' i
I V I I
I I I I
I
' I I
I I I I ! /1' I
I I
I i I I
/1 I I
I e
TS F!GLT;)E 3.12-5 8-9-73 I
I I/~
,/f* I i
I I
I i ! /i I I I I
I B*'\\f-"'.'0---
I I i I i I )" i
~..... I I
I 1/
' -l I
I i A i I
I.
I
/
I I
I I YI I I I I
I I
I I
YI i I i i
Vi i
I I
I l/
I I I I I I
I V
I I I I I
I I I I I I
I I I I I
I I I i I
I I I i !
I I
I o.o 0.2 0.4 0.6 FRACTIC:-~ FULL PO\\"/~R CHANGE ~O. 9
Q UJ
(
ti:
UJ en z
z 0 j::
.u
<C fE e
0.0 0.2 0.4 CHANGE NO.9,,
CONTROL B.A..~K IUSERT!ON LI!-UTS FOR 2 LOOP OPERATION WITH ONE INOPERABLE ROD i
V V I I I I 1,/,
I I
I I
I I l/1 i i I
I I
I i I /1 I I I I
I I
/
I I
i I I C\\t..';;{ D I
I/'
I I I i /:
I I
i I
), i I I I I
- /
I I I
I i I
0.6 a.a 1.0 0.0 I
I I I !
I I
I I
I. I I
I I
I I
I I
I I i
I I
I I.
l I i I i I ! I I
i I I I I I I I I
I I I
! I I I i
I I
i I I 0.2 0.4 FRACTION FULL PO\\"JER CH,'~;GE :._o. 9 i
I I
e 0.6 TS FIGURE 3.12-6 8-9-73
(
(
Unit l Amendment No. 35 Unit 2 Amendment No. 34 DELETE e
TS FIGURE 3.12-7 12-2-77 47
1.0 o.s 0.6 0
0 HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE SURRY POWER STATION
~TS FIGURE 3.12-8
-~~ ~~::+/-:=:::=~c==i:== *:=c'.:~=~ ::: t=~~ ::~:1::==::':J:c:: :::f:~:t-:::~<L =4-:::::Cc!::. :~:-
- §::=-=2.':=~;::=\\=~;=":+:: (:=' 1:'.=:t ::: -=d:::! ":Jc!: :;:t;;, ~,,f:'::'c~\\?::~: \\=~'.~ * =:*. *,~f;_~~li,;\\ ;:
2 4
6 8
10 12 CORE HEIGHT (FT.)
(
Amendment No. 51, Unit l Amendment No. 50, Unit 2 DELETED TS FIGURE 3.12-9 7-27-79 59
120 100 80 60
--~-
20 TS FIGURE 3.12-10 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED POWER SURRY POWER STATION
--+--,-
+-
- --t-*.
. --..==J:.:--=* -=i-
~-*-=r==~- ~-:t=*==t-= ::-
. *--r**-f---------+--
--+* - ~---+--
~._-__ ::
==l==l=:::i*l.:::
--~ -
i** -----
--t-- :_-:.-:i==--t--r--
I
-f-- --+
--f--*-r*
--r----
-~
- ------+-*
---1 F:.::.:i
=1= i-
. --=1--- -.
=-~ ---~ *-:-~t---1-.
- r---***
=-t==:
--~
- _ """--~ ----~=---=-- _ -;---- --~ -. -~-====p --~=:t- -~====¥---~~c-: -]----~
0..
_... ______________________... lall!im---*
-so
-40
-30
-20
-10 0
10 20 30 40 50 FLUX DIFFERENCE (AI)%
TS 6.6-9 The written report shall include, as a minimum, a completed copy of a licensee event report form.
Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.
(1)
Reactor protection system or engineering safety feature instrument settings which are found to be less conserv-ative than those established by the technical specifica-tions but which do not prevent the fulfillment of the functional requirements of affected systems.
(2)
Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.
Note:
Routine surveillance testing, instrument calibration, or preventative maintenance*which require system configura~ions as described in items 2.b(l) and 2.b(2) need not be reported except where test results themselves reveal a degraded mode as described above.
(3)
Observed inadequacies in the implementation of administra-tive or procedural controls which threaten to cause* reduc-tion of degree of redun~ancy provided in reactor protec-tion systems or engineered safety feature systems. *
(4)
Abnormal degradation of systems other than those specified in item 2.a(3) above designed to contain
e TS 3.12-l S 2.5 76 3.12 CONTROL ROD ASSEMBLIES AND POWER..DISTRllUTION*LIMI'!S Applicability es to the operation of the control rod assemblies and To ensure subcriticaJ.ity after a reactor trip, on potential reactivity inser ons from hypotheticaJ. conttol rod a Specification A.
Control Bank Insertion L
- l. Whenever the reactor is control rod assembly be fully withdrawn.
- 2.
When~er the l;'eactor
, except for physics tests and the shutdown control rods shall for physics tests and length control rod s 3.12-lA, 3.12-U, 3.12-3 for three-loop operation Figures 3.l2-4A,
- 3.
two-loop operati of physics calculations and physic obtained during unit startup and subsequent operation, in
/'
accordance wi.th the following:
- a.
The sequence of withdrawaJ. of the controllingb~, when going from zero to 100% power, is A, B, C, D.
- b.
An overlap of control banks, consistent with physics cal-Aaeadmc~ Ne, lQ
e TS 3.12-2 7 25 79 culations and physics data obtained durinr. Unit Startup and subsequent operation. will be permitted.
The shutdoYn r-:argin with allo*.. ~nce for a stuck control rod assembly be greater than or equal to 1.71% reactivity under all stea y-c::,.
~~
to full down incl.uciing effects of axial po*..1er distribution.
The shut-as used here is defined as the a~ount by whi n the reactor a
core subcritical at hot shutdown conditiocs T
.>547 F) no changes in xenon,
- 4.
\\Yhenever the reactor is rod position, i.e., the other reactivity zero po"Wer.
5
- Delet:ed
- 6.
Insertion limits do avg-that the highest and assu:ning critical criticality would be achia~ed in normal sequ~nce *ith ~=
than the inse=tion li:i: ::r during physics tests o during pe~icdic exercise of indivi
- rods.
However, the shutdown ind ica tad accve the low po"Wer physics rod wort:h and shutdo.n margin.
For this test the reactor may e critical full length control rod, expected to h~ve the h~~hest Amem:hneut rto. 50 UnH 1
.omeRGmeRt ~la, 49 YR it 2
- t.
i
- 7.
B.
e e
At all times except during low.power physics tation of 3.12.B.2.b.(2), the hot channel factors basis must meet the following limits:
TS. 3.12-3 11 26 i6
_______ __J
e Unit l Fq(Z) < 2.05/P x K(Z) for P > 0.5
. :l,,,,
FQ(Z) < 4.10 x K(Z) for P < 0.5.
N FAR~ l.55 (1 + 0.2(1-P)) x T(BU)
~ l.38/P
- Unit: 2 TS 3.12-4 5 16 80 Fq(Z) < 2.19/? x K(Z) far?> 0.5 Fq(Z) < 4.38 x K(Z) for P ~0.5
~H ~ 1.55 (l+o.2(1-P)) x T(BU)
~ ILOCA t.E. Assw *
. ~ ILOCA t.R Rod
- .:ha:re of rated power at whic:h is operating, K(Z) is 3.l2-8b for in TS Figure 3.l2-8a f Unit: l I and Figu=e the interilll given in TS Figure 3.12-9.
- 2. Prior to exceeding
~ach core loading, and during each e.f.fective full *power
- tion maps using the movable that the hot cham:ial factor For the purpose of ~his
- a.
The measurement o to *account ment the."effects of rod bow.
therea£~er, power distribu-shall be ma.de to confir.:.
speci£ication are satisfied.
Fijeas..
- shall be "increased tolerances~ measure-e measurement of enthalpy rise hot cbanne~ factor, the hot assembly e ~halpy rise factor, F~H!I.\\;.*,: and* the ho4 rod. en~hal~y* ~se facto. F!H[~~,.shall be_*
four percent to account
~r::ror. If _any channe1 factor ~ceeds its limit sp*e ux:.der 3.~2 *. B,l, the reactor power and. high neutron flux shall be reduced.until the li?llits under 3.12.B.l are met.
- hot channel factors cannot be brought to within ~he limits below wi.~hin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the Overpower~! and Overte:xpera~are ~T tTip se~points shall be similarly reduced.
the r
'(:'ni:: l
=~ < 2.05 x K(Z)
- /-= ~ l.55 J
- s "2.od
< 1.38
- <: 1.45
~----,
Uni:: Z F Q ~ 2.19 :< K(Z)
F:r::r
< 1 5-ilH -
-- J.l1-4a 5 Hi 80
//.
I
~ I LOCA. <1~47 IIBI AsS?::. -
N ILOwl F ~ Rod
~ *.S.::l AffleF1elffleAt r1e. *sa. Ul'lit 2
\\
e TS 3.12-5
§ '.9 79 (Unit 1)
- 3.
The reference equiJ.ibrium indicated axial flux difference (called
/
the target flux difference) at a given power level P0, is that,/*
indicated axial flux difference with the c:ore.in equilibri~~ncu conditions (small or no oscillation) and the control rods more than P/P0
- withdrawn.
The target flux difference at an~ other power P, is equal to the target value of P multipl~ d by the ratio, e target flux difference shall at least once per equiva be updated full power quarter.
The each effective difference must month of operation easurement, or by.lin ar interpolation using the most recent value a life.
- 4.
- Except as modified by axial *flux difference the target flux difference).
- a. At a power the value pred"cted for the end of the cycle a~ b, c, or d.below, the. indicated difference band about target: band* on a:dal flux of rated power, if in 15 minutes either restore its target indicated axial flux within the target band, or red e the reactor to less than 88 percent of rated power.
a power level no greater than 88 percent of rat (l)
The indicated axial flux difference may deviate from its target band.for a maximum of one hour (cumulative) in any 24-hour period provided the
- power, flux difference is within the limits shown on Figure 3.12-10.
Ameaciml!fl:£ Ne. 49, lffl:i.t l
(
TS 3.12-5
.s le 80 (Unit 2)
- 3.
I'he reference equilibrium indicated axial fll!.~ difference (called the target flu~ difference) at a given power level P0, is that I
/'
indicated a:-dal flux difference with the core in equilibrium. ~enon
/
no oscillation) and the control rods core than
/
withdra,rn..
!he target flux difference at a~6~her power level, P, is equal to the target value at P0 :::u.ltipli'ed by the*ratio,
?/P0 *
- ~gee flux ~i£ference shal.l. be measur~t le,.st once per e~-=.va.!.e:it _.,~ po-Jer quarter.
The carget flwi'difference must be
~y acf""n~1 :~aS"::"a:..:...
c,. or by li~ear inte elation using the most
=acSI:.: value a.=.d the alue predicted or the end of the cycle life.
- 4.
!:i:cept a.s :iodiiied by
~
fll.!:!: difference
, c, or d be!ow, the indicated intained within a +s: band *about tile t:arget. flux target band OU axial flu.~
diifere:ca).
- a. p.t a power: level ui.e indicated flux of rated power, if its target minutes either restore the indicated axial flux to
- b.
to within the target band, or educe the reactor power than 90 percellt of rated power.
power level no greater than 90 percent of ated power, The indicated axial flux differe~ce may dev1.a a f-rom it:s target band for a maxi:u:i of one hour (cu::iulative) in any 24-hour period provided the flu.~ difference is ~ithin the lil:d~s show-non Figure 12-10.
AnleAdmeRt ~lo. 58,. Unit 2
-~
One minute penalty is accumulated for each outside of the target band at power levels 50% of rated power.
IS 3.12-6 5 9 79 (Unit 1) one minute of operation' equal to or above
~
(2)
If 3.12.B.4.b(l) is violated, then the reactor power shaJ.l be reduced to less than 507. power within 30 minutes a (4) the high neutTon flux setpoint shall be reduced to o
gTeater than 557. power within the next four A power increase to a level greater than 88 power is contingent upon the indicated axial being within its target band.
testing of the Power be performed pursuant to of rated provided is maintained wit in the l.imits of Figure total of 16 be accumulated outside of the this testing
- c. At a power level no (1)
The indicated target band.
percent of rated power, difference may deviate from its (2)
- d. The rated power is c difference not One half greater than 50 percent of upon the indi~a;ed axial flux its target band for more the preceding is accumulated
.one minute of operation :ut:side of the target band rated power.
limits ations 3.12.B.4.a, be suspended during The power level is maintained at or below 85. of rated power 2 and The limits of Specification 3.12.B.l are mainta ned.
The power level shall be determined to be...!,_ 85%
f at least once per houT during physics tests.
that the l:!mits of Specification.:3.12.B.l are -being be demonstrated through in-core flux mapping at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
.-\\mencimene No. 49, *tfu:it l power shall-
(
(2)
(4)
- e e*
TS 3.12-6 S 16 80 (Unit 2)
Oce minute penal~y is accumulated for each one ainute of operation outside of the target band at power levels equal
/
to or above sor. of rated power.
/'
If 3.12.B.4.b(l) is violated, then the reactor power shall be. reduced to less than 50% power ~ithin 30 minutes /d the high neutron flux setpoint shall be reduced t no greater than 557. power w""i.thin the ne.~t four hou s.
A. pow"er.increase to a level greater than 90 p rc.e'l:l:t of rated power is contingent upon the i.Ildicated axi ~flux difference its target baud.
testi:lg of the Power be perfor.:ied pursuant is uia.btained of 16 of Neutron :'lu.x
!able 4.1-1 providad of Figure be accumulated this testing
- c. At a power level no (l)
T'~e indicated 50 percent of rated po~er, difference may d~viate from its t:arget band.
- d.
(2) great1:r th~u ~~- perci:t1.t of axial flux cliff erence its t~rget band for more ty during the preced;µig Que half minute penalty is accumulated levels bet-...-een 13%
li:nits c lila.Y be suspended during p ovided.:
of the target band ra:ted power.
- '!he:* po1-er level. is maintained at or belo~ 85%
£ rated power, and (2)
'!he lil:lits of Specification 3.12.B.l are mainta~e.
T"ne power level shall be deter=ined to be< as: of r ted power at least once per hour during physics cests.
Ve -i.fi-
.\\.
cacion chat Che lil:dts of met shall be demcusc:rated lease once per l2 hcu:rs.
Sp~c~ficatiou 3.12.3;1 are bei~
through in-core flux :!lapping at AmeRc:l~eRt ~le., * §S, UAi t 2
- ,_.,.. __ -- -- ----~ ---* --- *--
L e
TS 3.12-7
- 5 9 7~
Alarms shall normally be used to indicate the deviations from the axial flux difference requirements in 3.12.B.4.a and the flux difference time limits in 3.12.B.4.b and Co If the the The of service temporarily, the axial flux difference shall to the limits assessed, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and half-hourly thereafter.
alarm is hours after restoring the 5
- The allowable least once per 7 days whe once per hour for the operable status.
- 6. If, except testing, the quadrant to
- a.
- b.
- c.
average The hot and the power 3.12.B.1, or If the hot c e determined within 2 h0-urs et the specification of level and high neutron from rated power, 2% for trip setpoint to. average power tilt exceeds +/-10%, the power level and high neutron flux trip setpoint will o reduced from rated power, 2%
tilt.
AmeeElmcs.~ Ne, 49*, Yttie 1 AmenElmes.~ Ne, 48, tJaie 2
(
(
IS 3.12-8 11 26 76
- 7.
If, except for physics and rod exercise testing, after a further period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the power tilt in 3.12.B.S above is not
- 1.
rected to less than 2%:
- c.
If design hot channel factors for rated to the cause.of be made and reported as a reportabla clear Regulatory Commission.
greater than 10%, the uclear Regulatory Commissions power 11'! and percent for rated power be notified and th Nuclear Overpower, Over-
. ps shall be reduced one channel factor exceeds the the Nuclear Regulatory Commissio shall be reduced by assembly shall be considered inoper drive ~echanism, s misaligned* from its bank by more than 15 to average if the e assembly
-length control rod shall be considered inoperable if greater than l.S seconds to dashpot entry.
No mere than one inoperable control rod assembly shall be per-mitted when the reactor is critical.
- 3. If mere than. one control. rod ~ssembly in a given bank is out of service because of a single failure ex1:ernal to the indivi.dual rod drive mechanisms, i.e. programming circuitry, the provisions
,,._,.._.a ____ '"-
r
- 4.
TS 3.12-9 7-25-7~-
of 3.12.C.l and 3.12.C.2 shall not apply and the reactor may remain critical for a period not co e:-:c1:ed two hours provided In the event the.iffected assemblies cannot be returned within this specified period the reactor w* 1 be hot shutdown conditions.
of 3.12.C.l and 3.12.C.2 shall t apply during
- 5.
If an level which the assemblies are in ntionally misaligned.
full-length rod is loca d below the 200 seep of being tripped or if the full-length rod step level hether or not it is capable is of being tripped, then apply.
ion li:nits in IS Figura 3.12-:
- 6. If vn inoperable inoperable full-length be locacac, or if the and cannot be tripp a, then th~ inse tion limits in TS Fi;ura 3.12-3 apply.
- 7.
Deleted
- 8.
If a rod becomes inoperable and poten~ial ejected rod worth and opera.eiou peaking factors shall be by analysis within 30 days.
The analysis shall allowance for non-unifor::i fuel depletion in the of t:ie ino-perable rod.
If the analysis results in a more li::liti~g hypothetical transient than the cases reported in the safe:y analysis, the unit power level shall be reduced to an Ameru!ment Uo. 59, Unit 1 AffleFIEiffleAt Ne. 49, UHit 2
D.
E.
F.
be
- 1.
- 2.
The TS 3.12-10 7 2.§ 79 analytically determined part power.level which is consistent:
with t:he safety analysis.
out of service, the core day, and power level greater than lOi; or more than 30 rod motion.
be de ermined by one of the following methods:
- l. Movable detectors r quadrant)
- 2.
four per quadrant)
Ino
- l. If a rod position out of service then:
- a.
For operation position of RCC shall be (excore detector and/or~ vable incore detectors) to of rated pol;e-r, the core
/or ther.:ocouples subsequent
- rod, 24 steps, operation below 50% of rated power nos to.ring is required.
- ioni-more than one rod position indicator (~PI) channel nor t:wo R?I channels per bank shall be permitted to be at any :i:e.
~!isali~~ed o~ Jroooed Control Rod
- 1.
If t~~ loci ?osition Indicator Ch:1nnel is functional and the associated* full length control t":Jd is :nor~ t:han AmeASRleAt Ne. so ' Un; t l AroendmeRt Me. 49, Unit 2
TS 3.12-11 7 25 79 15 inches out: of alignment with its bank and cannot be realign d.,
then unless t:he hoc channel factors are shown to be limits as speciiicd in Section 3.12.B.1 within 8 as not: t:o exceed
- 2.
power above 75% of rat:ed power with a cont:rol rod more than 15 inches alignment wit:h an analysis shall first be ma to determine t:he hoc channel resulting al tvable power level Basis tration.
- and control of power
- s by the control groups.
during power oper ion ~ill place the reactor into concition.
The concrol assembly insertion limits provide hot shutdotm at.any time, assuming the highest worth con rod assembly sufficient margins to meet the accident analysis.
In addition, they provide a limit on erted ~od wor:h in the unlikely event of a hypothetical assembly and provide for acceptable nuclear peaking factors.
The limit may mined or. :~a basis of unit startup and operating data to provide a ~ora realistic li:it ~hich will allow for more flexibility. in unit oper~cion and Amefldmeflt Ne. SQ Amefldmeflt Ne. 49
, Uflit 1
, Uflit 2
(.
(
e TS 3.12-12 7 2S 79 still assure compli~nce wich Che shutdown require~ent.
The maxirnuc shut-down ::::irgin I"equir~ment occurs at end of co*re 1 i fe and is based on the the analysis of the hy?oCheci*cal steam break accident.
The li:its are based on end of core life condicions.
wich the excepcion of the che~ical control syst m malfunccion analysis are based on 1% reactivitY.
margin.
Relative positions control rod banks are determined a specified control rod bank overlap.
overlap is based on Che cons* eration of axial po~er shape control.
The specified control potential ejected rod worth revised to li~it tne to ace unt for the effects of fuel
\\ l densification.
The various control rod C and D) ar-e each t~ be moved the bank within one step (5/8 3,
i:i.
position. Position indication* is provided by tw methods:* a.digi al count of actuatii:g ?u.l.ses which shows the dem.:ind banks and linear position actual of the an Differential Transfor.:.er, ~hich indicates Transfor.ner is approximately.:!:. - of span steady state conditions.
The relative a indicator is such that, with the most adver if any t~o asse~~lies within a ~ank deviate
!n the event that the linear position indicator is service, the ef:ects of
.t\\meAameAt Ne. 59, ~n;t 1*
.t\\meAameRt ~le. 49, ijAit 2 of
TS 3.12-13 7 25 79
~alpositioned cont=ol roci asse~blies are cbservable from nuciear and proces/
/
in the }!ain Control Roct:1 and b:," core thern:ocouples enc:
Belew 5C:'7. pot,e't', no speci.:1.l monitoring :i.s quired
- ::,ositiot!eci control rod as:.ei.lb.lies with inoperable rod positio indicators (full cont-::-ol rod assembl7 12 feet out of aligr:JI;ent w:.th tion at sc:~
teady state power does not result in e:*:ceeding c
'Ihe specified. co th.it h.:i.'\\*e been asse:bly drop tir.e is bank) oper.i-safety analyses A.. -i inoperable control r assembly imposes additi.al dei:ia..'1.c!s on the operators.
TI:.e permissible number of assa:blies is lilti.:ed tc one in order to l~t turden, but such a fai:~re would not preyen.t dropping of trip.
Two criteria have been chosen.:1s fission gas release, pellet tempe First, the peak value of fuel out.rel rod asse!:"blies upc~ =eactor fuel psr=or.:a~ce rs:a:ed to
.:.ed:a:.::.ical properti:.s.
i::ust: cct e-:,::ceec !.'."::::"i.
Second,* the ~i:li.IJ:ur:i DNER operation or.in short tei-:::.
core must no be less than l.3C i~ n~r=~:
In addition tc the abo e, the peak ~inear power den ~ty, the nuclear c~=~a.l.py rise hot channel their lim.ting t:ett:peratu:e.
hot asse~bly entha!py which result from the large n the ECCS acceptance criteria l.i:r.it of f.ictor oust net==*::aec meet the initial cocdition assumed=~= t~e To aid in specifying the limits on factors are defi~ed.
AmeAEimeAt Ne. 69
- UR it 1 Amendment 610 49, UR it 2 f
-~
(
TS 3.12-14 5 Hi 80 FQ(Z), Height DeDendent Heat Flux Rot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by fuel rod heat flux, allowing for manufacturing tolerances on fuel sand rods.
Flux Hot Channel Factor, is allo-ws for local pellet density and diamete area statistically the surface heat flux.
F~, Nuclear Enthal the gap be~een is a factor of 1.03 to clad.
Combined to fuel rod Channel Factor,. is
'ned as the ratio of the integral of*1inear power alon power to the average rod power for both LOCA
_N ILOCA
~H Assm~' Hot Assembl Nuclear se Factor, is defined as t'he ratio
.of the integral of linear power power to the average assembly power It should be noted that the are used as such in the obtained by using take into account Tqus, the radial in radial (x-y) power at the point o*f o the enthalpy rise factors.
are based on integrals and Loca1 heat fluxes are licit power shapes which throughout the core.
is not necessarily the loss of -
criteria analyses are conservative.nth respect to specified in 10 CTR 50.46 using an upper bound acceptance
?.19 (Unit 2) times the hot channel factor nor.:alized given by !S Figures 3.12-Sa and 3.12-Sb.
Amendmerct No. 58, Unit 2
TS 3,12~15 7 27 79 When an FQ measurement is taken, measurement error, manufacturing tolerances, and the effects of r*od bo'i'1 must be allowed for.
Five percent is the app for measurement error for a full is the appropriate allowaµce for manufacturing toleranc
, and five to the In the tainties combined and result in a net increase pf that is applied uncer-normal operation of core is expected to result in ~H ~ 1,55 (l+ 0,2 (
the larger uncertainty ~
affecting FQ' (b) the operator ha a Birect influence on FQ through move~ent of rods, and can limit it to the red value, he has no direct control over ~a' and (c) an error tions for radial power shape, which may be detected during sta be compensated for by allowance for
(~40 thimbles system.
uncertainty for ~a influence FQ can*
is the appropriate from a full core map detector flux mapping The values specified for the limits of ~H,~~A andF!Hl~s are the values operation, AmeAS1ReAt Pie.
Am@Rdm@Rt Pie, It has been determined that four applied for measurement uncertainty Measurement of the hot channel factors physics tests, during each effective full power 51 s IJRit lo 5Q, l:JAit 2 as
and whenever abnormal. power clist:ribution conditions require a core power to a level. based on measured hot channel factors.
taken following core loading provides confi:ma.tion of the b~es including proper fuel loading patterns.
provides additioua.l assurance that identify operatioua.l anomalies which would, these bases.
trait l,\\meneme-ae Ne. 3S Unit 2 Amendmea: ~TG. 34 affect
(
TS 3.12-16 7 25 79 For nor:--..al operation, it has been determined that, provided certain c tions are obser-red, the enthalpy rise hot channel factor, c.e t; these conditions are as follows:
rod insertion differing by core than 13 inches/from the bar.k
/
- 2.
- 3.
demand position.
An indicated misalignment ~ii:iit of 13 steps p ecludes a rod misalignment no gr~ater t ~ 15 inches with cons"deration of ma:d.:num instrumentati
- error.
Control rod banks are sequenced banks as shown 3.12-lA, 3.12-lB, The ara not *:i~la.tad.
- 4.
Anal control flux difference control
!lank inse:--::.cn nux difference rs to the di!=arence top and bottom hal*,es of twc-se ion e:xcore neut:on dete~tors.
The flux difference is a measure o which is defined as tile difference in nor.::ali:ed the top_and bottom halves of the core.
pe==itted rela:i-.a.tion in ~H with decreasing p.ower power shape changes ~-iith red iI:.sertiou to t::.e i:lsert:!.on 1.:.,.,..: ts.
been ceter-~ned that provided the above conditions l through 4 are ti::.is hot cha::iel factor limit is met.
AmeRameAt ~le. 5Q, l:JA it 1 AmeRameRt Ne. 49 YRit 2
TS 3.12-16a 7 27 79 A recent evaluation of DNB test data obtained from experiments of fuel
//'
rod bowing in-thimble cells has identified that the reduction in DNBR due,/
thimble cells is mora than completely '.accommo.iated,;
thermal margins in the core design.
Therefore, it is not nec-N continue to apply a rod bow penalty to F~H*
Amei,elmeflt Ne. si,
AmeAe!meRt No. se UAit 1 l:lRit 2 I
\\
I
,/
08 §1 s'.
n-zrc si
)
cra.r.rna
-~
(.
TS 3.12-18
- 5 16 80 Ali!eRdriieRt Ne. 58, ldnit 2
- I I
I I
I I I
(
DEI.E'IED TS 3.12-19 5 1e ao e (~!).;.nd a reference-value w--hic~ corresponds to the ilibriwn value of axial offset (axial offset a ~I/fractional.ower).
reference value of flux di:ference varies t.lith p~er level.anc but e:q,ressed as axial offset it varies only ~ith bu...""nup.
Amenameflt Ne.. sa, llRit 2
**-*---* **----* -------~- *--*
- ---- --*--. **- ---------~*-*---*
(
TS 3.12-20 5 16 80 The technical specifications._on power distribution control given i.."'l 3.12.B.4 together w"'ith. the surveillance requirements g~ven in 3.12.B.2 assure that the Lit:liting Condition for Operation for the heat facto:- is c.et.
(or :reference) value of flux difference is
- tc.L:ows.
a:y t:.:::.e that equilibrium xenon conditions been estab-flux difference is noted with t full length rod steps witb.d:ra:.m (i.e. no al full power opera-as=--......,? prccee~s).
at.._;.,..: ch tb.e core "w-as fli.:z d.:.ffere:ce.
er=,or are necessary and the indicated reference fol:.~~g is required.
- core conditions :or
' for the ci:e in life~
u.a.1ly withdrawn farther divided by he fraction of f.ull power f 1 power value of the target ore power levels are obtained by fractional poYer.
Since the indi-allow--ances for excore detector tiou
- of +Si. AI are. pel:mitted frotll.
iods ~here e.~tensive load to establish the required ference every month.
FoT tilis reason, e specificatiou provides ttJo eds for *updating the St~ct c ntTol of the flux difference (and rod posi~iou) uring par~ po-er operation.
This is because xenon co trol at part power is not as significant as the control as ueces-P.me,:u:imeRt ~le *. a8, UF1i t 2
I r
I i
TS 3.12-21
.S 9 79 (Unit 1) power and allowance has been made in predicting the heat flu.~ peaking actors for less strict control at part power. Strict control erence is not aJ.ways possible during certain physics tests or duri 0 detector calibrations. Therefore, the specifications on pow restrictive during physics tests I
detector c "brations; this is acceptable due to the low signi-occurring during these operations.
In some cause rapid unit power reduction autom to deviate from the targ rod motion wi1.l when the reduced power level is reached.
tribution sufficiently to can be reached on a subsequent however period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> th.at the resulting affect the xenon dis-peaking factors which 11 power within the target band; a limitation of one hour in any outside the band.
This ensures significantly different e target. band.
The*. instan-insertion f:!."om those resul:ting from taneous consequences of limits are observed, worse than a 10 in peaking factor for the percent C;i:10.3 flux difference boundary is extende above, the essence of the procedure is to maintain in the core as close to the equilibrium full power r
e
.e
~
TS 3.12-21
.S 16 80 (Unit 2) po~er and allo~ance has been c.ade in predicting the heat flux peaking factors for less strict co~trol at part power.
Strict control of the fli.I,;t difference is net always possible during certain physics test~
calibrations.
Therefo_re, the specificat on
?c--er'\\_distri~utio~ control are less restrictive during physi e...-~re de.teeter c:a.li.brations; this is acceptable due to tests and probabili-
":J' oz: a accident occurring during these op
- S~2
= rauid ~'llit uo~er reduction rod motion will when the reduced c~*-:: :he !"fn-r pc-..:a= lavel is tti=~::ic::. s~ficiell::ly to c can ba *reached co.
be...:; l::.01'.;ever, to si:::plify the ar..7 pe:iod cf 24 e?:Sl!!"es ;that the insertion li~its a~e observed, the xenon dis-of peaking factors which*
full power within the target a limitztion of one hour in outside t~e band.
This distrib\\Jtious are net significantly within the target band.
the band, provided rod in ?ea.king fact.z'r for the* allowable flux difference percent increment 90% power, in the ra:ge z. 13.8,(perce!lt *{+10.8 percent indicated) --here 2 percent the permissible flux difference bounda e.xteuded above, the essence of the procedure is to maintain distribution in che care as close to che equilibriu:i full power ARl@l'lelffl@l'lt NtL 58 _ 111'1 He. 2
/
~
(
r e
e TS 3.12-22 11 26 76 I
as possible.
This is accomplished, by using the boron system to position the full length control rods to produce the required indicated flux i-erence.
A 2% quadrant near the core center sue al.lowauc:e.
control rods producing such where the maximum FQ occurs.
blaa.amene Nu * ?6 DEI.ETED present in the for disturbances rods and an error tilts up to 5% because m.isal.igned to the unrodded plane,
\\
Kt MO Ve TS Table 3.12-lA 5 9 79
.Ame:admeui: No 4 9-,--Ua:f..t 1meaEimeas We. 48, Yai~ 2
,/
(
REMOV~
e TS Table 3.12-1.B 5 16 80
.O.meRE':lmeAt Ne. es, UAit 2
I Rr:. MOVE TS Table 3.12-2 5 9 79 AmeeEimee:£ Ne, 49, Yftit l Amencimaat Ns. 48, Ua1t 2
1.0 0.8 N -O' r:.
Q ca l"
N I
~ s 0.6 i:::::
0 z I -
N -
0.4
~
0.2 C
0 2
HOT CHANNEL FACTOR NOR.""iALIZED SEE OPERATING ENVELOPE SURRY POWER STATION UNIT NO. l
-~---i=---
l 4
6 8
CORE HEIGHT ('FT.)
/
TS FIGURE 3.12-8a ~
§ lR go
-~--~:
t*--*~
==::-~- ----*
10 Amenelmeflt Ne. §8, UAi t 2
(
(
C N -
r:..O' Q
l:J
.N i 0 z I *-
N -
- ,=:
1.0 a.a
. 0~.6 0.4 0
?*
HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE SURRY POWER STATION UNI'! NO. 2 e
TS FIGURE 3.12-8):S
5 16 80
._:*-*= ---*
- ,~---
-=====:=-
......:-:::..-~---
- -1:::..--
- 1
- ~~1:: *****:-i:*-=--=.:.:.::.:;:
- ~7.:~J-:...** :-1:---;g;
-.i=-;
- .;:::==1:..* *:::i. ***--==:::! -
- -..:=~~:-- ~::.---:::;:---=-
4
.t
-~...
-~
---1ti.--..
6 8
CORE HEIGH'! (F'I.)
- = =
=
. y-=*==..111
.,-~
.'":\\.:-
=
- J!*
.;::-.. :.\\... -* -:
=---
=
10 AffieAeffleAt Pie. 68, f:IAit 2
1.0
- ~
- :.. !
0 0
2 HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE SURRY POWER STATION 4
6 8
CORE HEIGHT (FT.)
~TS FIGURE 3.12-8 10 12
{.
\\.
(
('
120 100 80 60 40 20 SEE P..,1T/.\\C.H ED PAG~
AXJ:AL FLUX DIFFERENCE LL'1ITS.
AS.:A FUNCTION OF R'I\\.TED POWER TS FIGURE 3.12-10 5 9 79 (Unit 1)
UNACCEPTABLE OPERATION trnActteT *.\\BiE ~
OPER..~ION
~~.
- ACCEP TION
(-29.5,50)
_(29.5,50)
-40
-30
-20
-10 0
10 20 30 40 FI.UX DIFFERENCE CAI) %
l
e TS FIGURE 3.12-10 u:,
- .:::c...**.
- - - -~.:....
~*-
AXIAL nux DIFFERENCE LIMITS AS A FUNCTION OF RATED POWER SUR.~Y POWER STATION
- I Q;
- l~
- =::: =
-* '==I
-. ::i:
- .... *c o*::..
N; SC
-~
40 20 4==tni.ACC':-:,!)~T~-
. :.-~ OP~.;.TIO~ ::--::,.-*
- -- *-*-*... i-
-+--* **
~--*-* -**-
- +
... --~
-50 30
-20
-10
--~-
0 10 nux DIFF!EENC~ (~I) %
'I,
.~~ UNACCEPT_.\\ELE=
- 'f= OPERATION
- .\\=____r-.
- 'i _j
~*---
20
-30 40 50 S 16 80-(Unil: 2)
AmeRameRt He. 58, URit 2 I.
I I I I
~
I I
. I I I I.
I I I.
120 100 80 40 20
-0 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED POWER SURRY POWER STATION TS FIGURE 3.12-10
- ~~ ;;~_;c.~;;:~ -~F--=~~: ~~~.~~~~-~'-=;.~.
- ~~+=== --
- 1 -~*--=~--==-+==-~
_:_~~~-~=:3:t--=---12=*==l===E==E=:E=:1
- --+*
- --+ --
-i-*
=--=t:==1*-=--1=1.+=*=1==*=1+-==f==-E-1--=-E=~=t:t*=*-==*.:~;=::-:_;_*
1--r-:--== -_,~= = *:=;
--~---*
--=r=
- - *---r
==i=- -
!----+--*-
- -r--
-~,___ --+-
-;-------t:-:r-=*=-::t:=:=t::::.::j:
=_-4-::t*==:
~ ~= -==""=*=-=---
t----f-- --+--_: --i:== -=--=+-==.:-:--.4--.. ---.-__
-r----i1-. =-~T=*:--:
- f _..
-50
-40
-30
-20
-10 0
10 20 30 40 50 FLUX DIFFERENCE (ti!)%
/
(
{
8*
TS 6.6-9 11 26 76
'Ihe written report shall include, as a minimum, a co~pleted copy of a licensee event report form.
Inforcation provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event.
(1)
Reactor protection system or engineered safety feature instrument settings which are found to be less conserv-ative than those established by the technical specifica-tions but which do not prevent the fulfill~ent oi the functional requirements of affected systems.
(2)
Conditions leading to operation in a degraded ~ode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.
Note:
Routine surveillance testing, ~nstrument calibration, or preventative maintenance which require system configurations as described in items 2.b(l) and 2.q(2) need not be reported except where test results themselves reveal a degraded mode as described above.
S~eei.:i.;.a.J.ly, tfte imfller-efteetieft ei 3.12.B.2.~.(2) is ftet re~er~aale.
(3)
Observed inadequacies in the imple~entation of administra-tion or procedural cont-rols which threaten to cause reduc-tion of degree of redundancy provided in reactor protec-tion systems or engineered safety feature systems.
(4)
Abnormal degradation of systems other than those-specified in item 2.a(3) above designed to contain