ML20076M414
| ML20076M414 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/29/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19268D962 | List: |
| References | |
| TAC-51054, NUDOCS 8307200361 | |
| Download: ML20076M414 (18) | |
Text
.
a r**(G y r
/*
9 UNITED STATES E.\\ ),;. c #i,
- i. NUCLEAR REGULATORY COMMIESION I*iW 2
WASHINGTtN, D. C. 20555 c
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 33 TO FACILITY OPERATING LICENSE NO. NPF-5 GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA EDWIN I..HATCU NUCLEAR PLANT, UNIT NO. 2 DOCKET NO. 50-366 1.0 In troduc tion By le t te r da ted Feb ruary ~2 3, L9.83.(Ra f.
1),, as supplemented by letter da ted April 19, 1983 (Ref. 2), Georgia Power Company (the licensee) has proposed changes to th e 'fechnical Specifications (TSs) for Hatch Unit 2 related to design co d if ic a tio n s that are being implemented at Hatch dnit 2 to reduce containment loads resulting from plant transients.
These TS changes would 1) lower th e opening and closing setpoints for actuation of four safety relief valves following initial actuation of any one of the four valves, and
- 2) lower the main steam isolation valve (MSIV) water level trip setpoint.
The February 23, 1983 letter encloses a document entitled "Edwin I. ' Hatch Plant Unit 2, Docket No.
50-366, Proposed Plant Modifications Low-Low Set Logic acd Lowered MSIV Water Level" (Ref. 3) which provides a detailed description of the proposed changes, a Safety Evalua tion of th es e changes, and proposed TS changes.
This docucent also includes, as appendices, two General Electric Company Reports NEDE 22223 (Ref. 4) and NEDE 22224 (Ref. 5) in support of the proposals.
h6 O
2-By letter dated March 30, 1983 (Ref. 6),as supplemented by letters dated May 10, 1983 (Ref. 7), May 20, 1983 (Ref. 8) and May 26, 1983 (Ref. 9), the licensee proposed TS changes that reflect changes to the core design for the third reload (Cycle 4) of Unit 2.
The March 30, 1983 letter encloses General Electric Company document Y1003J01A57 (Ref. 10) and General Electric Company letters numbers LMQ: 83-018 (Ref. 11) and LMQ: 83-022 (Ref. 12) in support of the proposed TS changes.
2.0 Evaluation 2.1 Low-Low Set Logic and Lovered MSIV Water Level System Response System Response The low-low set (LLS) relief logic modification for BWRs with Mark 1 containments is designed to prevent multiple subsequent actuations of saf ety relief valves (SRVs) which might normally b e expected during a transient f o llowing critical actuation of the SRVs.
This in turn will reduce or prevent the discharge loads on the containment and suppression pool structures resulting from subsequent SRV actuations.
The discharge loads from subsequent actuations tend to be higher due to the condensation of trapped steam in the safety relief valve discharge line (SRVDL) which results in a higher water leg in the SRVDL, and hence, larger thrust loads'on subsequent actuations.
In addition, the warmer steam air mixture in the SRVDL results in higher pressure air bubbles in the suppression pool, and the.refore, increased torus loads on subsequent actuations.
-wa v---
,. + -
,, _., -, ~,,,, _ - -,
.,,-e,m<-
g
. _ = _.
C The LLS for Hatch Unit 2 is an automatic SRV actuation system which, upon initiation, will assign preset opening and closing setpoints to four preselected SRVs.
These setpoints are selected such that the LLS controlled SRVs will stay open longer, thus releasing more steam (energy) to the suppression pool, and hence more energy (and time) will be required for repressurization and subsequent SRV openings.
The LLS increases the time between (or prevents) subsequent actuations sufficiently to allow the high water leg created from the initial SRV opening to return to (or fall below) its normal water level, thus, reducing thrust loads from subsequent actuations to within their design limits.
In addition, since the LLS is designed to. limit SRV subsequent actuations to one valve, torus loads will also be reduced.
The lower MSIV water level trip causes the MSIV closure actuation to be changed from a reactor water level 2 signal to a reactor water level 1 signal.
This design modification maintains the main condenser availability for a longer time which allows more energy to be released to the main condenser and will result in a slower repressurization rate.
The lower MSIV water level trip reduces isolations, SRV challenges and provides some benefit to SRV subsequent actuations.
The TS changes requested by the licensee reflect these logic modifications to 1) lower the opening and closing setpoints for actua tion of the four selected SRVs following initial
+-
e
- vag,
. g vy -
,g,q, y-%.
.,w,-*i__
~
.... + -.
1,
6 e
actuation of any one of the four and 2) to lower the MSIV water level trip from level 2 to level 1.
We have reviewed the licensee's submittal as discussed above and find that: 1) the design modifications are compatible with normal operations and other safety systems, 2) the licensee has demonstrated by analysis of limiting transienes that the LLS will extend SRV subsequent actuation time sufficiently to clear the water column in the SRV discharge line, and 3) the design modifications will not adversely affect the plant performance or safety margins.
LLS Circuitry Design The LLS circuitry consists of four redundant logic channels, each of which actuates one SRV.
There are eleven SRVs at Hatch Unit 2, seven of which are actuated by the Automatic Depressurization System (ADS).
The four non-ADS SRVs will be used for the LLS function.
Each of the four LLS controlled SRVs will open when their respective solenoid becomes energized by the LLS logic.
The LLS Logic channels that actuate SRVs 2B21-F0138 and 2B21-F013F, channels A and C respectively, are powered by 25 Vdc from division 1 Class 1E supply 2H11-P925.
LLS logic channels B and D (SRVs 2821-F013G and 2821-F0130 respectively) are supplied from division 2 Class 1E supply 2H11-P926.
.n e--
-q,w
- y4y,
--,,y-m-
w--
g-
---v-w er v
e-y-.
--v.
w-
Es 1
~.6
In order f or an LLS channel to energize its solenoid, both an arming logic and an initiation logic must be satisfied.
The arming logic is satisfied when any SRV has opened and reactor pressure has exceeded the high pressure scram setpoint (this setpoint is selected above the reactor protection system high reactor pressure scram setting to assure that a scram has occurred).
Four separate reactor pressure instrument channels (one for each LLS channel), each consisting of a transmitter and associated trip unit, have been added to provide this reactor high pressure permissive function in the LLS arming logic.
Each transmitter and trip unit are powered from the same division as their corresponding logic channels.
Once the arming logic for any LLS channel is satisfied, it is sealed in and annunciated in the control room, and remains sealed in until manually reset by the operator.
In addition, the arming logic in either LLS channel of the same division will seat in the arming logic in the remaining LLS channel of that division provided the reactor high pressure permissive in that channel is satisfied.
Initial SRV actuations are detected by two sets of pressure switches located in the SRV discharge lines.
Each discharge line contains one pressure switch powered from division 1 and the other from division 2.
Contacts from these switches are used in the arming logic of the corresponding divisional LLS Logics.
These pressure switches are set above the nornal pressure expected in the discharge line (85 psig).
- Once armed, the LLS actuation / control logie uses newly added reactor pressure instrumentation to control the LLS SRV sole-noids, thus opening and closing these SRVs at their assigned LLS setpoints.
The actuation / control logic remains in effect as long as the arming logic is sealed in.
The added instrumen-tation consists of one transmitter and an associated trip unit for each of the four LLS logic channels.
In addition, a second trip unit associated with the transmitter providing the arming logic pressure p e r m,i s s i v e for each LLS channel has been added and is used in the actuation / control logic for that channel.
Both trip units providing' control for a given SRV have the same setpoints such that they actuate simultaneously.
This arrange-ment prevents single failures within the transmitter and trip unit portion of the LLS circuitry from causing a spurious SRV opening once the arming logic is satisfied, and from causing a SRV to remain open after reactor pressure has decreased to the reclose setpoint.
The added transmitters and trip units are powered from the same division as their corresponding logic channels.
All four LLS logic channels can be tested at power.
Test status lights in the control room indicate when the arming logic relays and contacts have operated satisfactorily during testing.
These I
l test lights can also be used to verify proper operation of the seal-in and reset circuits.
Each LLS channel provides annunciation i
-: r ----
'e
. in the control room upon loss of power.
Test switches are provided to verify operability of this power monitor function.
Additional test lights in the control room are used to verify operability of the trip units used in the LLS actuation / control logic during testing.
The proposed Hatch Unit 2 TS changes associated with the LLS modification call for monthly channel functional tests of all reactor pressure instru-ment channels (used for both the arming logic permissive and SRV control / actuation).
A channel functional test of all SRV discharge line pressure switches will also be performed monthly (portions of these channels inside the primary containment may be excluded from this test).
Channel calibrations and LLS logic system functional tests will be performed during each refueling outage.
This test frequency is consistent with the test interval for the ADS and is acceptable.
The LLS circuitry contains no channel or operating bypasses.
The circuitry added for the LLS function is located in the con-trol room and is separated in accordance with IEEE 384-1974 The components of the LLS system (including power supplies) are classified as Class 1E.
The LLS will remain operable in.the event of loss of offsite power.
LLS components located inside the drywell are qualified for the environmental conditions associated with a small break LOCA.
- w -~.w g
O O
Based on our review as discussed above, we have determined that the LLS modification installed at Hatch Unit 2 is designed to perform its intended function given a single failure.
In addition, no single f ailure in the electrical circuits could be found which would cause more than one SRV to stick open.
The LLS is designed in accordance with the requirements of IEEE Standard 273-1971.
Conclusions On the basis of our review and findings as discussed above, we conclude that the design modifications and the proposed TS changes related to the LLS relief logic and lowered MSIV water level trip are ac c e p t ab le.
2.2 Cycle 4 Reload The reload application involves (1) the replacement of 236 spent fuel assemblies with fresh P8x8R fuel assemblies and some p,
riously irradiated 8x8R, (2) the analysis of safety cons id era tions involved in the determination of Cycle 4 operating limits, and (3) the proposed modification of existing TSs and the addition of new TSs to reflect the changes made in the composition of the core for Cycle 4.
Fuel System Design The Hatch Unit 2 Cycle 4 core will contain 560 fuel assemblies of which 236 will be changed during the current Cycle 4 outage.
About midway through Cycle 3 operation, fuel failures became evident from increasing coolant activity.
General Electric Company (GE) believes that the failure mechanism was crud-induced localized corrosion (CILC) inasmuch as the Hatch Unit 2 Cycle 3 core contained batches of Zircaloy cladding believed r
g-n
..-pe-.
_9_
to be susceptible to this type of degradation.
To limit the number of failures, the licensee restricted the power level to 70% of rated power for the latter part of cycle 3.
With assistance from GE, the licensee anticipated (using a GE proprietary waterside-corrosion fuel-failure model) that the i
Hatch 2 failures would be in the initial-core assemblies and consequently structured the Cycle 4 fuel management scheme accordingly.
However, the Cycle 4 outage sipping and visual examination revealed failures in 19 Reload-1 assemblies.
Therefore, the licensee has recently revised (Ref. 7) the Cycle 4 fuel loading scheme that was origin'aT1y described in the Supplemental Reload Licensing Submittal (Ref. 10).
Since fuel failures should not be an expected occurrence during an operating cycle and since compliance with TS activity limits does not assure acceptable fuel performance, we asked the licensee to provide some assurance that additional'CILC failures would not occur during Cycle 4.
The licensee responded (Ref. 8) that all rods from the affected lots / ingot were either discharged or passed a visual and NDT examination and, therefore, that 9
no CILC failures-were expected in. Cycle 4.
We thus conclude that the licensee has provided reasonable assurance that cycle
-[
4 will be operated without a recurrence of CILC failures.
The 236 replacement assemblies to be installed for Cycle 4 operation will be fresh P8X8R assemblies and previously irra-diated assemblies taken from Unit 1.
The Unit i replacement assemblies will consist of reconstituted Reload-2 (i.e.,
8X8R design) assemblies.
The Cycle 4 core composition is summarized in Table 1.
The Cycle 4 fuel assemblies are thus standard designs that have been' previously approved for application
-y%_wew-eww w,-e,vw w
y--e-
7v*w 1r-tuf-'--ww wv-
--'v' er
--'--'-'--V9
-7N'-w
" - - - * - ~ ~
---'---vi-r* - - * - - " "
- D--
-m*T
=
---e w
4,
10-in the Hatch Units; consequently, the mechanical aspects of the reload fuel require no further NRC review.
Table 1 HATCH-2 CORE INVENTORY Assembly Designation Cycle Loaded Number 8DRB221 (IC) 1 108 P80RB284LA 2
96 P8DRB283 3
120 P80RB265H (fresh) 4 132 80RB265H (irradiated) 4 104 560 The licensee's analysis of other considerations involved in the determination of Cycle 4 operating limits is presented in the reload safety analysis (Ref. 10).
In all fuel-design-related areas except those separately identified, the reload report relies on the generic report, General Electric :3tandard Application for Reactor Fuel (Ref. 13).
Reference 13 has been reviewed and approved by the NRC staff.
We conclude that additional staff review o'f those portions of Reference 13concerning the standard fuel design is unnecessary for the Cycle 4 application.
The licensee's submittal provided both new and revised MAPLHGR limits.
The new MAPLHGR limits are for the fresh fuel (P8X8R).
w'-
-7 e
---7e,-
yg-
-s wmy 3
,w y
7 m
-r,
--my---,,w-we wo y
r, - -gyw
. The revised MAPLHGR limits.for all fuel types have been extended to accommodate exposures to as much as 45 GWd/ST.
These limits were generated by methods previously approved (Ref. 14).
Although the methodology used is generically applicable for the MAPLHGR limit determination, we believe that the effects of enhanced fission gas release in high-burnup (i.e., greater than 20 GWd/MTU) were not adequately considered in the fuel performance model.
In response to this concern, GE requested (Refs. 15 and 16) that credit for approved, but unapplied, ECCS evaluation model -changes and calculated peak cladding temperature margin be used to avoid MAPLHGR penalties at higher burnups.
This proposal was found accep table (Ref. 17) provided that certain plant-speciffc conditions were met.
The licensee has stated (Ref. 7) that the GE proposal is applicable to the Hatch Unit 2 ana. lysis with the exception of three exposure points where the calcu-l lated peak cladding temperatures for Hatch Unit 2 fuel exceed those assumed in the GE analysis.
In those three cases, the peak cladding temperature for the Hatch Unit 2 fuel is less 1
I than 200F la;Ser ch an the limiting temperature in the generic analysis.
O,3 censee has also stated that a temperature change of less than 20 F (per 10CFR50, Appendix K, Section II.1.b) is not considered significant.
We accept this conclusion and conclude that the MAPLHGR Limits proposed for 1
Cycle 4 operation of Hatch Unit 2 are acceptable.
l 1
i l
~
- ~ ^
. g. - T.G
T a.
Thermal and Hydraulic Design A sa f e ty limit.MCPR has been 4.mposed to assure than 99.9.
percent of the fuel rods in the core are not expected to experience boiling transition during normal and anticipated operational transients.
A,s stated in Reference 13, the approved safety limit MCPR is 1.07.
The safety limit MCPR of 1.07 is used for Hatch 2 Cycle 4 operation.
The most limiting events have been analyzed by the licensee to determine which event could potentially induce the largest reduction in the initial critical power ratio (a CPR).
The a,CPR values given in Section 9 of Reference 10 are plant specific valaes calculated by the methods including ODYN Methods.
The calculated ACPRs are adjusted to reflect either Option A or Option B ACPRs by employing the conversion methods d es crib ed in Re f e renc e 18.
The MCPR values are determined by adding the adjus ted aCPRs to the safety limit MCPR.
Section 11 of Ref er en ce 10 presents both the cycle MCPR values of the pressurization and non-pressurization transients.
The maximum cycle MCPR values (Options A and B) in Section 11 l
are specified as the operating limit MCPRs and incorporated
(
into the TSs.
For the MCPR limiting event, feedwater controller failure to maximum demand, the analysis has l
assumed operation of the high water level (Level 8) trip and the turbine bypass systems.
We informed the licensee by the letter dated May 12, 1983 (Ref. 19) that since operation of the Level 8 trip and the turbine bypass system are assumed l
e 0.
in the analysis, and neither of these systems have been demonstrated to be qualified to operate (i.e.,
not saf ety grade), TSs are required to assure their operability.
The licensee's response (Ref. 9) disagreed with the need for these TSs and indicated that they would like to discuss thiserequirement 'further prior to committing to permanent TSs.
However, they submitted proposed TSs for the Level 8 trip and the turbine bypass systams as we requested so that these TSs would be available to us for inclusion in the reload amend-wr if w= diti wo t-ert low attditietTri~ time- -f or disetmriotr trf the -ta ct er and required implemention of these ISs for the start of Cycle 4 oper-ations.
We believe that we should allow the licensee more time to present their arguments and discuss this request, and based on our review of the significance of these trip systems with regard to limiting transients during the fuel cycle, we have decided to defer implementation of these TSs for 60 days following startup in order to allow time for further discussion of this subject with the licensee.
Since the approved method was used to determine the operating limit MCPRs to avoid violation of the safety limit MCPR in the event of any anticipated transients, we conclude that these limits are ac c e p tab le.
The licensee has proposed the use of the MCPR limits currently in the Hatch Unit 2 TSs (Figures 3.2.3-1 and 3.2.3-2) for Cycle I operation at increased core flow up to 1057. of rated i
flow.
We find that the operating MCPRs based on the reload
}
analyses for Cycle 4 (Ref. 10) are lower than the calculated values for the Cycle 3 core, and on this basis condlude that the use of the currently approved operating limit MCPRs in T
,.~.r
, - ~ - - -. -
.--.-----,-.m..,-.,,..-..m.-.-
.--r, - ~,, - - -
s 6
. Figures 3.2.3-1 and 3.2.3-2 of the TSs are acceptable for Cycle 4 operation.
The results of thermal-hydraulic analysis (Ref. 10) show that maximum reactor core stability decay ratio is about 0.83, which is same as the calculated value for the Cycle 3 core which has been.previously approved.
We therefore conclude that the thermal-hydraulic stability results are acceptable for Cycle 4 operation.
We find that approved thermal-hydraulic methods have been used and that results of analyses support the proposed limit MCPRs, which avoid violation of the safety limit MCPR for design transients.
Nuclear Design The Hatch Unit 2 Cycle 4 reload will consist of 560 fuel bundles as shown in Table 1.
The initial core loading had a maximum average enrichment of 1.87 w/o in U-235.
The reload fuel is similar in physical design to the initial core load fuel, but it has a maximum average. ennichment of 2.84 w/o in U-23.*
All fuel bundles consist of 62 fuel rods and 2 water rods.
The active fuel length is 150 inches.
The shutdown margin of the new core meets the TS requirement that the core be at least.25%AK suberitical in the worst reactive condition when the highest worth control rod is fully withdrawn and all other rods are fully inserted.
For Hatch Unit 2 Cycle 4 GE calculated that the k,gg under cold conditions and the strongest rod out is equal to.985 resulting in a shutdown margin of 3.3%4K.
--~
w
,,-~--,,y w,
+,a-y
-1;-
The standby liquid control system is capable of bringing the reactor from full power to a cold shutdown condition assuming none of the withdrawn control rods is inserted.
The 600 ppm boron concentration will bring the reactor suberitical to k,ff =.950 at 20 C xenon free conditions (Ref. 10).
Based on our review of the licensee's submittal (Ref. 6) and the plant specific analysis (Ref. 10), we have determined that the nuclear characteristics and the expected performance of the reload core for Ratch Unit 2 Cycle 4 are acceptable.
Conclusions On the basis of our review of the reload safety analysis for Cycle 4 op eration o f Hatch Unit 2 including the proposed related changes to the Hatch Unit 2 TSs, as discussed above, we con-clude that this core reload will not adversely affect the capability to operate Hatch Unit 2 safely during Cycle 4 operation and that the proposed related TS changes are acceptable.
3.0 Environmental Considerations l
We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact i
and. pursuant to 10 CFR 951.5(d)(4), that an environmental i
~
p
- impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of 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 this amendment will not be inimical to the common defense and security or to the health and safety of the public.
4 y
e
-1 7 -
6.0 References 1.
Letter, Georgia Power Company to Director of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, dated February 23, 1983.
2.
Letter, Georgia Power Company to Director of Nuclear Reactor Regulation, U.S.
Nuclear Regulatory Commission, dated April 19, 1983.
3.
Document entitled "Edwin I.
Hatch Plant Unit 2, Docket 50-366, Proposed Plant Modifications - Low-Low Set Logic and MSIV Water Level", provided as an Enclosure to Reference 1.
4.
General Electric Company Report " Low-Low Set Logic and Lower MSIV Water Level Trip for BWRs with Mark I Containment,"
NEDE 22223, September 1982. Proprietary.
5.
General Electric Company Report " Low-Low Set Relief Logic System and Lower MSIV Water Level Trip for Edwin 1. Hatch Nuclear Plants Units 1 and 2," NEDE 22224, December 1982.
Proprietary.
6.
Letter, Georgia Power Company to Director of Nuclear Reactor Regulation, U.S.
Nuclear Regulatory Commission, dated' March 30, 1983.
7.
Letter, Georgia Power Company to Director of Nuclear Reactor l
Regulation, U.S.
Nuclear Regulatory Commission, dated May 10, 1983.
8.
Letter, Georgia Power Company to Director of Nuclear Reactor Regulation, U.S.
Nuclear Regulatory Commission, dated May 20, 1983.
9.
Letter, Georgia Power Company to Director of Nuclear Reactor Regulation, U.S.
Nuclear. Regulatory Commission dated May 26, 1983.
10.
General Electric Company Document " Supplemental' Reload Licensing Submittal for Edwin I.
Hatch Nuclear Plant Unit 2 Reload 3 (Cycle 4), Y1003J01A57, January 1983.
11.
Letter, General Electric Company, L.K.
Mathews, Southern Company Services, dated Pebruary 22, 1983.
Letter No.
LMQ-83-018.
12.
Letter, General Electric Company to R.
D.
Baker, Georgia Power Company dated February 24, 1983, Letter No. LMQ: 83-022.
Y o
13.
" General Electric Standard Application for Reactor Fuel",
GE Report NEDE-2400ll-P-A-4, January 1982.
14.
D.
G.
Eisenhut (NRC) letter to E.
D.
Fuller (GE), June 30, 1977.
15.
R.
E.
Engel (GE) letter to T.
A.
Ippolito (NRC), May 6, 1981.
16.
R.
E.
Engel (GE) letter to T.
A.
Ippolito (NRC), May 28, 1981.
17.
L.
S. Rubenstein (NRC) memorandum for T.
Novak, " Extension of General Electric Emergency Core Cooling Systems Performance Limits," June 25, 1981.
18.
" Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," GE Report NEDE-24154-P, October 1978.
19.
Letter, John Stolz (NRC) to J.
- Beckham, Jr.,
Georgia Power Company, dated May 12, 1983.
Dated: June 29,1983 Principal Reviewers:
K.
Desai R.
Kendall D.
Powers S.
Sun L.
Lois l
l
'~
~. _ _, _ -