ML20151Y432
| ML20151Y432 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 01/17/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20151Y426 | List: |
| References | |
| TAC-59309, NUDOCS 8602120695 | |
| Download: ML20151Y432 (9) | |
Text
1 g>2rraq UNITED STATES
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!"3,- I,j NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REG SUPPORTI:4G AMEN 0 MENT NO.121 TO FACILITY OPERATING LICENSE NO. DPR-57 GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 DOCKET NO. 50-321 k
1.0 INTRODUCTION
24, 1985 (NED-85-528), Georgia Power Company (GPC)
By letter dated July proposed an amendment to the Hatch Nuclear Plant Unit 1 (HNP-1) Technical Numerous change items are identified in the submittal that Specifications.
support the installation of the analog transmitter trip syst # (ATTS).
The installation of the ATTS was previously reviewed and approved by the NRC in Amendment 103 to the HNP-1 Operating License.
The ATTS is a new design for portions of the system instrumentation of the It was Reactor Protective Systems (RPS) of Boiling Water Reactor (BWRs).
developed by the General Electric Company (GE) and is being supplied as original equipment in later built BWRs (e.g., BWR 6). The design was adapted GE developed the ATTS to offset operating to the HNP-1 as a backfit.
disadvantages of the digital sensor switches of the original safety system instrumentation.
The principal objective of the ATTS is to improve sensor intelligence and reliability while enhancing testing procedures.
2.0 EVALUATION Nomenclature Changes to the Technical Specifications The ATTS modification replaces pressure, level, and temperature digital switches in the RPS with analog / trip unit combinations. The digital switches are identified in the instrument description of the current Technical The licensee proposes to change the instruments listed in the Specification.
This Technical Specification to reflect the installation of the new ATTS.
change is acceptable to the staff.
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Modification of the Surveillance Frequency The licensee proposes that the surveillance frequencies for the ATTS equipment be changed from that listed in the existing Technical Specification The licensee proposes the following (for equipment replaced by the ATTS).
surveillance frequencies:
Once per shift for channel check Once per month for channel functional test Once per operating cycle for channel calibration The channel check once per shif t is a new requirement for the ATTS Such a test was not applicable for the mechanical switches equipment.
The addition of this requirement is, cherefore, a replaced by the ATTS.
change toward the more conservative surveillance for the ATTS equipment.
The channel functional test required each month is either as conservative or more conservative than required by the existing Technical Specifications.
The channel calibration frequency of once per operating cycle is less conservative than the present HNP-1 requirement for calibrations which in However, currently approved Technical most cases is once every 3 months.
Specifications for BWR6's, which utilize the ATTS system, provide channel calibrations based on a frequency of once per operating cycle. Once per operating cycle channel calibration frequencies have been previously approved for other ATTS instrumentation for Hatch Unit 1 (e.g., HNP-1 Operating License Amendment 103).
The GE report NEDE-22154-1 (the supporting GE document for the installation of the ATTS in Hatch Unit 1) recommends transmitter calibration onceIt per operating cycle when the reactor is out of service for refu design
<hich can be between 12 and 18 months.
The primary factor in setting the calibration surveillance frequency is the drift of the transmitters and trip units. The total loop accuracy and the Setpoint drift is total loop drift are added to obtain the trip setpoint.
the only value that is extrapolated in the licensee's setpoint methodology.
In many cases, the manufacturer's specifications only provide drift values i
for 6 to 12 month intervals. These values were extrapolated linearly to j!
provide 18 to 24 month drift values for use in the Hatch setpoint
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The licensee intends to evaluate the performance of the ATTS against the h
manufacturer's specifications and, if necessary, propose modifications to the The surveillance frequencies specified in the Technical Specifications.
Definition of Surveillance Frequency states that the operating cycle interval as pertaining to instrument and electrical surveillance shall never exceed 15 Therefore, the proposed requirement for calibration once per months.
operating cycle is a requirement to calibrate once per operating cycle or l
i Current Standard once per 15 months, whichever is the shorter interval.
j Technical Specifications require calibration once per 18 months.
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Based on the above information, the staff finds that Technical Specification changes requiring channel checks of the ATTS equipment once per shift and 3
channel fuactional tests once per month of the ATTS e i
l months are acceptable.
The reactor vessel water level, shroud water level and reactor pressure post-accident monitoring instruments all receive input from ATT l
instruments.
Two new being replaced with qualified class IE devices compatible with ATTS.
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recorders are also being added.
The licensee proposed that the calibration frequency be changed for these I
instruments to once every operating cycle except that the recorders b j
calibrated once per 12 months.for the recorders is once per 12 months for l
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years for the indicators.
I The staff finds that calibration of the recorders each 12 mo acceptable.each operating cycle (which as previously discussed in the section on The staff i
modification of surveillance frequency cannot exceed 15 months).
finds that calibration of this equipment once per operating cycle not to 4
j exceed 15 months is acceptable.
4 Deletion of Drywell Pressure Sensors E11-N011A, B, C, D l
The original design of HNP-1 has the high drywell pressure signa i
l coming from eight sensing devices.
MPL numbers) provide signals to reactor heat removal (RHR), core spray (C high pressure core injection (HPCI) systems; E11-N010A, B, C, D (existin numbers) provide signals to the automatic depressurization system (ADS).
This configuration is inconsistent with the inputs for reactor water levels 1
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i and 2 which are provided by only four sensing devices, namely B21-NO31A, The licensee proposes to,make drywell pressure 3
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B, C, D (existing MPL numbers). sensor configuration consistent with w l
l, drywell pressure sensors E11-N010A, B, C, D to provide s i
i The reliability of the drywell pressure trip logic for ECCS will not be affecte systems of ECCS.
J Plant safety margin is not being reduced since the
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level of sensor redundancy for each trip function is maintained.
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This change deletes instruments E11-N011A, 8, C, D and transfers their t
Since these associated trip function to instruments E11-N010A, B, C, D.
j irstruments (E11-N010A, B, C, D) are being incorporated into the ATTS mudification, the instrument number was changed to E11-N694A, B, C, D.
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j The proposed Technical Specifications revision changes the Remarks i
3.2-4, 3.2-5, and 3.2-6 to include all the functions of dryw j
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sensors E11-N694A, 8, C, D.
this modification and the proposed Technical Specification change, as i
l discussed above is acceptable.
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, 4 RCIC Turbine Exhaust Pressure Trip Setpoint Modifications The proposed trip setpoint value for the reactor core isolation cooling (RCIC) turbine exhaust pressure is 45 psig compared to 25 psig in t l
The objective of this change is to increase the Technical Specifications.
RCIC system availability for some small and intermed Company report NEDC-30136 which was provided with the submittal.
In Hatch Unit 1, as in other BWR plants, the exhaust line pressure sign The used to trip the RCIC turbine since it could indicate line blockage.
high pressure signal from the space between the two rupture diaphra exhaust line is used to initiate closure of the isolation val i,
For Hatch Unit 1, both pressure signals are j
turbine steam supply line.
However, only the diaphragm included in the Technical Specifications.
pressure signal is included in the Standard Technical Specifications l
Technical Specifications of other BWR/4 plants that were checked during t l
review.
The justification for the proposed increase in NEDC-30136 included considerations of the beneficial effects of increased availability as well as drawbacks such as increase in offsite and onsite doses resulting from the r
l higher leak rates from the turbine gland seals and from governor and stop The proposed increase in valve stems at the higher exhaust pressures.
j setpoint would not affect normal system operation or the consequences l
large LOCAs (for the LOCA, rapid system blowdown would prevent RCIC However, the proposed increase could be beneficial for some jl small and intermediate break LOCAs involving significant increases in operations).
containment pressure when the RCIC system could provide an alternate j,
of makeup water and prevent fuel damage. increase permits longer
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f The higher signal as the result of the increase in containment pressure i
l are well below 10CFR20 limits.
- j On the basis of our review of the justification in NE
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ll reactor safety and is acceptable.
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Trip Setpoint/ Allowable Values For Rosemount Transmitters
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The proposed change involves revision of the trip setpoint/ allowable v 1
l for reactor vessel levels 1, 2 and 3, shroud water level and reactor _ ves 3
It was stated that a) the trip j
j steam dome pressure low instruments.
setpoint/ allowable values were calculated with me l
i proposed values are more conservative with respect to the analytic
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We requested and reviewed the General Electric l
then the present values.
Company justifications for the analytical limits.
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The reactor water level 1 signal initiates closure of the MSIVs, startup of the Core Spray Spray system and the RHR system in the LPCI mod of the ADS initiation signals.12 inches above the top of the active fuel) is t K calculations (ref.1).
The reactor vessel level 2 signal initiates HPCI, RCIC and the SBGT systems.
It is also used for initiating isolation of secondary containment andThe p partially isolating primary containment.
The General Electric Company in the Appendix K calculations was -38 inches.
conducted a sensitivity study to evaluate the effect of changing this value and found that the use of a new value of -58 inches for level 2 had Hence, a effect on the analyzed accident / transient consequences" (ref. 1).
new value of -58 inches was proposed in the submittal to reduce unnecessary challenges to the HPCI, RCIC and isolation systems.
The reactor vessel water level 3 provides one of the ADS initiation signals The analytical limit of +1.5 and partially isolates primary containment. inches is the value used in t The reactor shroud water level 0 signal is used as an interlock to prevent In reference 1 it is stated diversion of LPCI flow to containment spray.-
that the analytical limit of -211 inches meets the specification of.Section 7.4.3.5.4 of the Hatch Unit 1 FSAR that this interlock be set at a value i
lower than two-thirds core height.
The reactor vessel steam dome pressure signal for the recirculation pump discharge valves has an analytical limit of 300 psig to conform to theThe Appendix K calculation assumptions.
33 seconds after the pressure drops to this value (ref. 1).
On the basis of our review of the proposed setpoints and the justification provided in reference 1, we conclude that the proposed changes are accepta l
Reactor Vessel Steam Dome Pressure Permissive Modifications fo Injection valves
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_The CS and LPCI injection valves have both upper bound and lower bound limits
.l' The upper for the reactor vessel steam dome pressure permissive signals.
bound limit helps provide overpressurization protection for these low l
pressure systems and has a value of 500 psig in the current Technical f '
The lower limit is the pressure at which the valves are I
Specifications.
In this submittal, it is i
assumed to start opening in the LOCA analyses.This proposal was made because the i
proposed to eliminate the upper limit. difference between the current l
small to permit meeting the limits with the current setpoint methodology and We have reviewed the the specifications of the new pressure transmitters.
i proposed change and justification and conclude that the current upper boun l
limit, which helps prevent overpressurization of the CS and RHR should be retained.
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lower limit with a resulting small increase in the calculated peak cladWe und temperature (PCT) for the limiting break. evaluating the expected s another submittal.
Miscellaneous Trip Setpoint/ Allowable Value Modifications Modifications to the trip setpoint allowable values were proposed for 25 These are listed in Table 1.
In reference RPS and ECCS trip functions.
2 and the present submittal, the licensee, in response to staff questions, stated that a) unless noted as such in the submittal, the analytical limits used in the setpoint calculations were the original limits used in the Hatch Unit I safety analyses, b) any changes to the limits had been justified by a safety evaluation and c) "in no case with these new limits do the FSAR analyzed transients or accidents exceed the safety limits which are spec in the Plant Hatch Technical Specifications".
The bases for the analytical limits were audited and discussions of particular limits associated with the RHR, CS, RCIC and HPCI systems were We conclude that the proposed miscellaneous held with the licensee.
modifications are acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
An Environmental Assessment and Finding of No Significant Impact has been issued for this amendment.
4.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations, and the issuance of these amendments will not be inimical to the comm defense and security or to the health and safety of the public.
5.0 REFERENCES
Letter from L. T. Gucwa, Georgia Power, to Director of NRR,
" Response to Staff Questions on Proposed Technical Specification l
1.
Changes for ATTS," December 7,1985.
Letter from L. T. Gucwa, Georgia Power, to J. F. Stolz, NRC, i
2.
June 7, 1984.
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Dated: January 17, 1986 Principal Contributors:
J. Mauck and C. Graves
Attachment:
Table 1
q TABLE 1 Miscellanerus S.tpoint Modifications RPS Trip Function ECCS Trip Function Maip steam line flow - high Drywell pressure - high RHR pump discharge pressure - high Main steam line tunnel RHR pump flow - low temperature - high Core spray pump discharge pressure - high Core spray pump discharge Reactor vessel steam dome flow - low pressure - low permissive HPCI steam supply Drywell pressure - high pressure - low RWCU area temperature - high HPCI pump discharge flow - high, low RWCU area ventilation HPCI pump suction differential temperature - high pressure - low HPCI turbine exhaust diaphragm pressure - high Suppression chamber water level - high HPCI turbine exhaust pressure - high HPCI emergency area cooler
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ambient temperature - high i
RCIC pump discharge flow - high, low I
RCIC pump suction pressure - low i
RCIC pump suction
!i pressure - low RCIC steam supply pressure - low IeIk RCIC turbine exhaust diaphragem pressure - high RCIC steam line differential pressure - high I,
Suppression chamber ambient temperature - high Suppression chamber differential temperature - high j
I RCIC emergency area cooler ambient temperature - high
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7590-01 O
U.S. NUCLEAR REGULATORY COMMISSION P
GEORGIA POWER COMPANY, ET AL NOTICE OF ISSUANCE OF AMEN 0 MENT TO FACILITY OPERATING LICENSE The U. S. Nuclear Regulatory Commission (the Commission) has is No.121 to Facility Operating License No. DPR-57, issued to Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority fo Georgia, City of Dalton, Georgia (the licensees), which revised the T Specifications (TSs) for operation of the Edwin I. Hatch Nuclear Pla The amendment is No.1 (the facility) located in Appling County, Georgia.
effective as of the date of its issuance and shall be implemented within 30 days.
This amendment revises the TSs for Hatch Unit I to support the instal It includes changes to the of the analog transmitter trip system (ATTS).
surveillance frequencies and trip setpoints associated with the ATTS j
equipment.
The application for the amendment complies with the standards and l
requirements of the Atomic Energy Act of 1954, as amended (the Act),
The Commission has made appropriate Commission's rules and regulations.
i findings as required by the Act and the Commission's rules and regula 10 CFR Chapter I, which are set forth in the license amendment.
Notice of Consideration of Issuance of Amendment and Opportunity for P i
Hearing in connection with this action was published in the FEDERAL i
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No request for a hearing or petition for leave
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on August 26, 1985 (50FR 34559).
to intervene was filed following this notice.
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7590-01 4
o Also, in connection with this action, the Commission prepared an Environmental Assessment and Finding of No Significant Impact which was 9, 1986 (51 FR 1051).
published in the FEDERAL REGISTER on January For further details with respect to this action, see (1) the application for amendment dated July 24,1985,(2) Amendment No. 1 21to License No.
All of these DPR-57, and (3) the Commission's related Safety Evaluation.
items are available for public inspection at the Commission's Public 20555, and at the Appling Document Room,1717 H. Street, N.W., Washington, D.C.
31513.
A copy County Public Library, 301 City Hall Drive, Baxley, Georgia of items (2) and (3) may be obtained upon request addressed to the U.S.
20555, Attention: Director, Nuclear Regulatory Commission, Washington, D.C Division of Licensing.
Dated at Bethesda, Maryland, this 17th day of January 1986.
F THE NUCLEAR REGULATORY COMMISSION f.f Daniel R. Muller, Director
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Project Directorate #2 Division of BWR Licensing i
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