ML19309H337
| ML19309H337 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/17/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19309H335 | List: |
| References | |
| NUDOCS 8005130072 | |
| Download: ML19309H337 (7) | |
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C' SAFETY EVALUATION 3Y THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING Af1ENDMENT N9. 73 TO FACILITY OPERATING LICENSE NO. DPR-57 AND AMENDMENT N0.14 TO FACILITY OPEPATING LICENSE NO. NPF-5 GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION fiUNICD/1 ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA EDWIN 1. FATCH NUCLEAR PLANT UNIT NOS.1 & 2 DOCKE~i NOS. 50-321 AND 50-366 A.
Average Power Range Monit.or ( APRM) Rod Block and Scram Setooints I.
INTRODUCTION By letter dated October 18, 1978, Georgia Power Company (licensee) requested an amendren-to the Technical Specifications appended to Operating License ! a. D?R-57 for the Edwin I. Hatch Nuclear Plant Unit No. 1.
The proposed amndment would (1) substitute aquivalent terninology for corou:ation of Average Power Range Monitor ( APRM)
Rod Block and Scrar setooints and (2) revise associated surveillance requirenents.
By letter dated June 5,1979, the licensee requested a similar amend-ment to the Technical Specifications appended to Operating License No.
NPF-6 for the Edwin I. Hatch Nuclear Plant Unit No. 2.
This amendment was requested to provide for consistent procedural usage between both units.
II.
EVALUATION (1) APRH Rod Block / Scram Setpoints The current Technical Specifications require adjustment of the APRM rod block and scram setpoints in the event of operation with the maximum total peaking factor (MTPF) exceeding the design total peaking factor {DTPF). Under such conditions the setpoint is adjusted by the ratio DTPF/MTPF to ensure that the fuel cladding integrity limits are not exceeded during anticipated operational transients.
The Hatch I core currently contains three types of fuel (7x7, 8xE and 8x8R) each with different DTPFs. Therefore, three formulas are required to ensure compliance with the stated specification.
< The licensee proposed to substitute the equivalent expression FRP/CMFLPD for DTPF/MTPF, where:
FRP is the fraction of rated power CMFLPD is the Core Maximum Fraction of limiting power densi'cy.
Using.this terminology, a single formula with a unique solution is obtained.
In the Unit 2 core, the assemblies are all the same length and contain the same number of rods. The principle reason for making the change for Unit 2 is to provide for consistent procedural usage between units.
The staff has previously reviewed and approved (Reference 1) the 1
proposed tenninology as being equivalent. The equivalency is explained as follows The DTPF can be expressed as the design linear heat generation rate divided by the plant rated thermal power per unit length of fuel rod.
In a similar manner the MTPF can be expressed as the maximum linear heat generation rate divided by the plant operating power per unit length of fuel rod. The CMFLPD is defined as the highest value of the ratio of the linear heat generation rate existing at a given location to the design linear heat generation rate for the bundle type. From these definitions it is easily determined that the ratio DTPF/MTPF is the ratio of the design linear heat generation rate to the maximum linear heat generation rates times the fraction of rated thermal power, or 1/CMFLPD times FRP. Thus FRP/CMFLPD and DTPF/MTPF are equivalent.
Currently, APRM rod block and trip settings are adjusted through multiplication by the ratio of DTPF/MTPF. Such a reduction in set points is required in the event of operation with MTPF>DTPF.
Instead of multiplying the APRM set points by FRP/CMFLPD the same result can be achieved by multiplying the APRM reading by CMFLPD/FRP to get a gain-adjusted APRM reading.
If the reactor is operating in a st'eady state mode the APRM reading (before gain adjustment) is equal to FRP. Therefore by adjusting the gain until the APRM reading is equal to CMFLPD, the APRM reading has effectively been multiplied by CMFLPD/FRP as required.
(2) APRM Indicated Power Level During normal operation the actual core peaking factors can exceed the design values. This variance is accommode;ted by reducing the APRf1 setpoints to retain the same desired " margin to trip." The
" margin to trip" can be reduced either by lowering the trip set-point or by increasing the APRM indicated power level.
From the viewpoint of equipment performance, these two methods are equivalent.
o From the viewpoint of reactor operator performance, causing a power monitor to indicate a level that is different from the true power level' introduces an additional source of possible operator confusion.
IEEE Standard 279 Section 4.20 specifies the design principle that indications that could be confusing to the reactor operator should be minimized. Because the APRM setpoints are flow-biased, recali-bration involves re-adjusting several parameters with external test equipment. Our review of the calibration procedure and dis-cussion with operating personnel indicates that re-calibration requires about one hour for each of the six APRM channels. The design does not lend itself to a simple re-calibration procedure.
The present APRM instrumentation channels do have gain adjustments to adjust the channel response toward a more accurate indication with respect to true power as determined by a heat balance.
Routine use of the gain adjustment to cause the channels to read higher than actual is not consistent with the general design principles of IEEE 279, and could lead to operation in a non-conservative manner.
When full power and temperature equilibrium are attained, setpoint re-adjustments are very infrequent. During a reactor startup and approach to full power, the APRM setpoints may need to be re-adjusted several times. To require a lengthy six-hour calibration procedure several times during approach to full power is not reasonable nor in the best interest of safety.
We have determined that use of the APRM gain adjustments to maintain an adequate " margin to trip" during full power operation is not a justifiable deviation from the general principles.
However, within certain limitations, use of APRM gain adjustment can be allowed during an approach to full power. The limitations appropriate for allowing the gain to be used to maintain the APRM
" margin to trip" are as follows:
(1) gain adjustment should be used only when the reactor is less than 95% of rated power; (2) the magnitude of such adjustments should be less than 10%
of rated po'wer not to exceed 100% full power indicated; (3) any intentional inaccuracy of the APRM channels should be made obvious to the reactor operations staff at all times. Appropriate log entries are to be made for each such gain adjustment.
Each affected APRM indicator should be marked in an obvious manner to identify the offset between true power and the power level indicated by the APRM instrumentation.
. (3) Surveillance Requirements The licensee proposed that the CMFLPD be determined daily during reactor power operation equal to or greater than 25%. The daily frequency for determining the CMFLPD is identical with the current specification for determining MTPF. However, the revised sur-veillance requirement would be applicable only above a specified power level. W2 have reviewed the licensee's submittal and deter-mined that a power level of 25% above which the CMFLPD would be determined is acceptable. This is acceptable because below 25%
power the ratio of the peak LHGR to core average LHGR would have to exceed 12 (for 8x8 fuel) in order for LGHR to be at limiting val ue. Such large peaking factors are highly improbable because permissible rod patterns preclude such large peaking factors. Based on the above and current licensing practices as set forth in Ref. 2, the licensee's proposal is acceptable.
III.
CONCLUSIONS We have determined that deviation from the prescribed APRM calibration procedures during full power operations is not consistent with the general principles of IEEE 279. However, during reactor startup and approach to full power, the APRM channel gain adjustments may be used, within limitations, to change the " margin to trip " The limitations involved are enumerated above.
To summarize, in the event of operation between 25% and 95% power with CMFLPD greater than FRP, the APRM gain shall be adjusted such that the APRM reading >100% times CMFLPD. This change does not involve a reduction in margin to Die trip point and prescribes a more direct use of limits monitoring data from the plant process computer.
In addition adjusting the APRM gain is much easier than changing the APRM trip setting, so that there is less chance for human error.
For operation above 95% power, with CMFLPD greater than FRP, APPfi rod block and trip settings will continue to be adjusted by multiplication by the ratio of FRP/CMFLPD. The proposed Technical Specifications have been modified to be consistent with the above restraints. The necessary i
revisions were discussed with and agreed to by the licensee.
B.
Oglethorpe Power Corporation Name Change I.
INTRODUCTION By letter dated November 8,1979 the licensee proposed an amendment to Operating Licenses DPR-57 and NPF-5, consisting of a name change in the operating licenses.
. II. EVALUATION On November 7,1978, Oglethorpe Electric Membership Corporation changed its name to the Oglethorpe Power Corporation (OPC). Because OPC is a co-owner of Hatch Plant Units 1 and 2, it is necessary to amend the licenses to correctly identify the co-owners.
III. CONCLUSION The change of name of a co-owner is an administrative activ.. = nct is appropriate at this time. Therefore the licenses will be modii ed to correctly identify the co-owners in this amendment.
C.
Hatch Unit 2 Thermal Power Monitor I.
INTRODUCTION By letter dated February 28, 1980 the licensee proposed a change to the Plant Hatch Unit 2 Technical Specifications to modify the Average Power Range Monitor (APFM) high-high flux scram trip logic. The proposed amendment would replace the present APRM trip logic with circuitry to condition the APRM flux through a first order low pass filter that has a 6 second RC time constant. This circuit represents the fuel dynamics which'will approxinate the reactor thermal power during a transient or steady state condition. The intent of this modification is to avoid spurious scrams caused by momentary anomalous neutron flux spikes.
This nodification was previously licensed on Plant Hatch Unit No.1 by Amendment No. 69 to Facility Operating License No. DPR-57.
II. EVALUATION (1) Thermal Power Monitor The Thermal Power Monitor (TPM), also called an APRM Simulated Thermal Power (STP) Trip in some documents, is a modification to the APRM trip system. The modified system generates two trips:
a trip with a flow biased setpoint and a second trip with a set-point fixed at 120% power. The flow biased setpoint is unchanged from that presently in the Technical Specifications. However, the TPM conditions the APRM output to apply a time constant of about six seconds, which is less than but comparable to the fuel thermal time constants (seven to ten seconds). Thus, the signal compared to the fi)w biased setpoint is a conservative simulation of fuel rod heat flux. This feature allows the plant operator to avoid spurious trips due to minor neutron flux overshoots when maneuvering the reactor.
4
. The signal which is compared to the fixed 120% power setpoint is not modified. Thus, there is always a " fast scram" at 120% power in addition to tha " heat flux" scram which may be below 120% power, depending on flow.
(2)
Effect of TPM on Safety Analyshs Since all the transient analyses are done assuming full design flow, the TPM has no effect because the 120% " fast" trip is identical to the original system at full power. Any effect due to the TPM must be on analyses which are initiated from low flow conditions.
GE has addressed the analysis of the various transients initiated from low flow conditions on pp. 5-8 of Reference 3.
The generic analyses described there show that only the idle recirculation pump startup, recirculation flow controller failure (increasing),
feedwater controller failure (max demand), and rod withdrawal error can become more severe at low flow conditions. This is the basis for the flow-dependent multiplier (K ) in every GE plant's
~
f Technical Specifications.
The analyses supporting the Ky factor did not take credit for the flow biasing, but instead emnservatively assumed the trip to occur at 120% power. Therefore, the analyses supporting the flow-dependent multiplier (K ) remain bounding.
f Similarly, the various accident analyses which involve a neutron flux induced trip (e.g., rod drop accident) assume the trip to occur at 120'(,' power regardless of initial power or flow conditions.
Therefore, the validity of the nccident analyses is also unaffected by the introduction of the TPM.
III. ENVIRONMENTAL CONSIDERATION We have determined that the amendments do not authorize a change in effluent types o'r total amounts not an increase in power level and will not result.in any significant environmental impact.
Having'made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement, negative declaration, or environmental impact appraisal need not be prepared in connection with the issuance of the amendments.
IV. CONCLUSION We have concluded, based on the considerations discussed above, that:
1 (1) because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable
N assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendments will not be inimical to the conmon defense and security or to the health and safety of the public.
V.
REFERENCES (1) Safety Evaluation by NRR Supporting Amendment No. 35 to DPR-33, January 10, 1978, Docket No. 50-259 (2) NUREG-0123 Rev.1, Standard Technical Specifications for General Electric Boiling Water Reactors, April 1,1978 (3) General Electric Boiling Water Reactor Generic Reload Application, NEDE-240ll-P-A, May 1979 Dated:
April 17, 1980
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