ML20009F330
| ML20009F330 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 07/22/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20009F328 | List: |
| References | |
| NUDOCS 8107300301 | |
| Download: ML20009F330 (69) | |
Text
.
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E NUCLEAR REGULATORY COMMISSION hJ j
.usmNoron. o. c. 20sss t Mys/ J s,.....f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 69 FACILITY OPERATING LICENSE NO. DPR-3 YANKEE ATOMIC ELECTRIC COMPANY YANKEE NUCLEAR POWER STATION (YANKEE R0WE)
DOCKET NO. 50-29 INTRODUCTION By letter dated March 26, 1981 (as supplemented May 27, 1981, and July 8, 1981), Yankee Atomic Electric Company (the licensee). requested changes to the Technical Specifications for the Yankee Nuclear Power Station (Yankee Rowe). The - cpose of the changes is to permit the Yankee Rowe reactor to operate with a reloaded core (Core XV). The-proposed changes also permit operation with certain changes to the Main Steam Non-Return Valves and the Auxiliary Feedwater (AFW) system.
DISCUSSION The Yankee Rowe reactor core consists of 76 fuel assemblies, each having a 16 x.J array of fuel rods. The Core XV reloaded core will be loaded with a two region configuration of 36 fresh and 40 recycled assemblies fabri-cated by Exxon Nuclear Company. The licensee provided an extensive descrip-tion of the replacement fuel in reports transmitted by letters dated July 4, 1975, and November 7,1975, and we approved the use of the fuel in the Yankee Rowe reactor in License Amendment No. 21 dated December 4, 1975.
8107300301 810722 PDR ADOCK 05000029 P
o The licensee has provided, and we have reviewed, nuclear and thermal evaluations and transient and accident analyses with the reloaded core.
LOCA analyses are based on use of a revised lower plenum 1hase separation model (YAEC 1231, Rev. 1) submitted on December 10, 1980, and approved by the staff on June 19, 1981.
A.
EVALUATION OF NUCLEAR DESIGN, CONTROL ROD WITHDRAWAL TRANSIENT, R00 DROP TRANSIENT, R0D CJECTION ACCIDEkT, AND PHYSICS-RELATED TECHNICAL SPECIFICATIONS - YANKEE R0WE 1.
Introduction The parforr.anse analysis of Core XV for the Yaatcc-Roue olant is der.cribed in report YAEC-1240 entitled "Yantee ';uclear Peaer Station, Core XV Perfor.rance Analysis." The nuclear design description and the analyses of the control rod withdra.eal transient, the rod drop transient, and the rod ejection accident have been reviewed.
In addition the proposed changes to the physics-related Technical Specifications have been re-viewed.
2.
Evaluation The reload configuration is described and the nuclear parameters of Core
- .V are corpared to those of the preceeding Core XIV. The fresh fuel to be lcaded on the periphery of the core has a 3.5 weight percent enrich-2nt ccrpared to a 4.0 percent enrichment for Core XIV. As a result, "Le critical boron cencer.tration at begir.ning of cycle (B00) is sr:11er for Core XV.
This leads to slight differences in nuclear rirareters, especially the BOC moderator temperature and void coefficients whi:h are i s 2 regatd ve for Core X".
TS.e saricus piru ~eters are calculated iy techr,icues previously erric;.ed by Yankee Atomic and the differences fr:m
- cre XIV are consistent with the char.ge in cycle design.
On this !2 sis i.e find the discussion e' r.uclear design acceptible.
'le ha/e reviewed the general discussion of transis t and accident input conditicos. '?e conclude that t!.e reference cycle alues bcunu the values for Core XV c/ cept for core inlet temperatste and oparating lim,t DNBR. C;eration of the core witi, Core XV inlet t3x;;rature values has been previously apprcved for Corcs XIII and XIV ted is acceptable here.
The reactivity coefficients for Core XV are bounded by those for the
.ererence cycle and are acceptable.
The ef fect :f the sr311 decrease in 0;2 rating limit DNBR is discusscd sepurately for 53ch event for which it is impcrtant.
The ef#ect of the reduced operating DNSR limit en the contrcl rod withdrawal transient is shown to be minimal.
The n.inicum DNBR for this transient is greater than 2.0 compared to a safety limit of 1.3.
Similarly, the results of the rod drop incident sre orily mildly affected.
If cradit is taken fcr the r=daced total pEsling #ictor for Core XV then
.?e cc ::: ences of :he incident are less savere :ran those far the raferer:e cycle.
Litewise the important input para eters for the rod efecticn eccident are bounded by the mest recent analysis of the event, that fte C;re XIV.
Iased on the ab:ve dit:sssio-de #ind the accidant analyses for Core XV to be acce; table for these events.
3.
Conclusion _
We have reviewed the proposed Technical Specification changes regarding nuclear physics parameters. On the basis that previously employed methods were used to derive the specifications we find the following changes to be l
acceptable:
l l
l
1 Specification 3/4.1.1 - Shutdown Margin Figure 3.1 Control Rod Insertion Limits Figure 3.2 Factor F as a function of Rod Insertion Figure 3.2 Core XV Multiples for Xenan Redistribution Figure 3.2 Core XV Multiples for Reduces Power Specification 5.3.1 - Fuel Assembly Design Features e
. B.
YAtlKEE R0WE CYCLE XV RELOAD SER (PLAT 1T TRAf451EllT AND ACCIDEf4T AtlALYSES) e 1.
Boron Dilution An inadvertent boron dilution event adds positive reactivity to the core by reducing the boron concentration in the primary coolant. This produces power and temperature increases in the reactor core and may cause loss of shutdown margin, depending on the core conditions at the time of the dilution event.
The baron dilution event was reanalyzed for Yankee Rowe Cycle XV reload to provide the complete spectrum of boron dilution analyses covering all possible plant operational modes (Ibdes 1 through 6) as currently defined in the Technical Specification. The analyses are performed assuming the most limiting combination of initial boron concentration, inverse boron worth, reactor coolant system volume, reactor coolant system temperature and the dilution rate for each plant operating condition.
For Modes 4, 5 and 6 the licensee assumed a minimum reactor coolant system volume of 3
1276 f t, which conservatively accounts for possible upper vessel head -
draining and main coolant loop isolation. Technical Specificatidn limits l
on shutdown margin (-4.72%Ap), and Keff (0.99) are also assumed in the analyses. The Technical Specifications are being modified to require Keff <.96 for operation in Modes 4 and 5.
A Keff <.96 will allow withdrawal of 1% ap in control rods while maintaining a shutdown margin of 5% ap during operation in Modes 4 and 5.
The licensee has stated that there are various indications and alarms (e.g., high level alann on CVCS low pressure surge tank, high RCS average temperature and high core power alams for Mode 1 and 2 operation, high flux recorder alarm, High level, temperature and pressure alarms in the pressur-izer surge tank for Mode 3 operation, high flux level recorder alam, high level, temperature and pressere in the CVCS low pressure surge tank
. for Mode 4 and 5 operation, and high neutron flux level recorder alarm for Mode 6 operation) which would indicate the boron dilution events to plant operators.
In addition, audible alarms on two independent source range and two independent intermediate range channels, all powered from the plant vital bus, would provide indication of rising neutron count rates in case of a boron dilution accident. Based on our review of the licensee's submittal, we conclude that the licensee has demonstrated that the plant is adequately protected for postulated boron dilution events in all modes assuming the worst single active failure.
2.
Isolated Loop Startup Incident If a reactor coolant loop is isolated from the primary coolant system and subsequently brought back into this system without first matching the boron concentration and temperature of the isolated loop to those of the system, an increase in core reactivity and power may occur. Such an event could lead to DIB and possible fuel damage unless suitable pro-cedures and protection are provided.
In order to provide additional protection against the occurrence of an isolated loop startup event, interlocks are placed on the cold leg isolation valve controls so that they cannot be opened unless the. isolated loop temperature is within 30*F of the hottest active loop cold leg temperature.
The licensee has provided in a letter dated 3/29/74 (Ref 9-5) the results of an analysis for the Cycle XI reload to demonstrate that the i
reactor protective system precludes fuel 6amage for the worst possible
temperature mismatch between the ' active and isolated loops. The result indicated that the minimum DNBR was greater than 2.97 and the peak fuel centerline temperature was less than 3485*F during the transient and was acceptable.
The licensee has stated ir. the core XV performance analysis (Ref. 9-6) that the most significant parameter that influences this event is the value of the coderator temperature coefficient of reactivity (MTC). The more negative the coefficient, the greater the reactivity insertion during the event. The limiting case for the reference cycle is at end of life, when the MTC is most negative. For Core XV, the MTC is always less negative than the value used in the Core XI limiting case, so the results will always be bounded by the reference cycle limiting case which is acceptable.
1 3.
Loss of Load Event A rapid and large reduction in power demand on the rcactor during full power operation results in a corresponding reduction in the rate of heat removal from the primary coolant system.
Such an event could lead to system overpressurization and subsequent core damage if suitable pro-tection were not provided.
The licensee has provided the results of a bounding analysis in Reference 9-S for the Cycle XI reload using the design value of the :aoderator temper-l ature coefficient'at B0C. The licenses also provided additional information
9 in a letter dated November 21,1978 (Ref. 9-7) to demonstrate that the moderator temperature coefficient and doppler coefficient have a minor impact on the results of the loss of load transient.
For the Cycle XV reload, the parametric studies provided in Reference 9-7 conservatively bound core XV partmeters. Hence, it is concluded that the core XV plant response to a loss of load event would be within system pressure and fuel design limits and therefore acceptable.
l 4.
Loss of Feedwater Flow Event A rapid and large decrease in feedwater flow when operating at power without a corresponding reduction in steam flow would lead to a decrease of water inventory in the steam generators.
In the event of a total loss of main feedwater, an auxiliary feedwater system is available to provide water to the steam generators and the reactor protection system responds automatically to trip the plant.
The most recent analysis of the loss of feedwater flow event was performed during the cycle XIV operation and submitted in the licensee's letter dated September 12,1979, (Ref. 9-8). The analysis provided in the reference letter concluded that the plant response to this event was acceptable.
Since Cycle XV operating conditions are completely bounded by the analysis presented in reference 9-8, it is concluded that the Cycle XV perfunuance to a loss of feedwater event is acceptable.
5.
Loss of Coolant Flow Event l
l A loss of coolant flow could result from a mechanical or. electrical failure i
l in one or more of the reactor coolant pumps. The immediate effect of the l
l
reduction in coolant flow is an increase in :oolant temperature. This increase in temperature could lead to DNB and subsequent fuel damage if suitable protection were not provided.
The licensee has provided the re.sults of an analysis in Reference 9-5 which demonstrates that the Cycle XI perfomance with a loss of coolant flow event was acceptable.
The above analysis also demonstrated that the loss of coolant flow trans. nt is sensitive to core parameters, reactor protection system setpoints and steady state thermal margin.
In each of these areas, except steady state thermal margin, the Cycle XV parameters are bounded by the reference cycle (Cycle XI) parameters.
At steady state design conditions, the fresh fuel DNBR for the reload core is approximately 3 percent less than the reference core. Addition-ally, although the reload core physics data are more favorable, the Technical Specification limit on moderator temperature coefficient of
.2 x 10 -4ap/*F is not bounded by the reference analysis.
The licensee has perfomed an analysis of a complete loss of coolant flow during four loop operation to detemine the impact of the above two parameter differences. The results of this reanalysis confimed that the reference cycle analysis continues to bound Cycle XV plant operation and therefore it is acceptable.
6.
Steam Line Rupture Accident The licensee's last complete main steam line rupture analysis was per-fomed in 1973 in support of the Cycle XI reload submittal and submitted to the NRC via Reference 9-5.
Subsequent cycles, including Cycle XIV, were licensed via reference to this Cycle XI analysis because the refuel-ing changes 1) did not impact the themal hydraulic transient and 2) the care physics parameters were bounded by the core XI analysis.
E
.. In the Cycle XI reload analyses, the circumferential rupture of the main steam header with the most negative moderator teirperature coefficient was analyzed for different operating conditions. There is no return to power following the reactor trip and the critical heat flux is not exceeded.
Therefore no DNB occurs and fuel temperatures remain below 4700*F. During the steam line rupture accident, the reactor coolant system and steam gen-erator pressure and the radioactive release are within acceptable limits.
The analysis assumes that the most reactive control rod is stuck in the withdrawn position. The licensee reported that a loss of offsite power was considered for this analysis, but the most limiting case should be when offsite power is available to drive the reactor coolant pumps resulting in a more rapid cooldown.
The steam line rupture accident addressed in the' Cycle XV reload analysis referred to the most recent investigation into a steam line rupture at Yankee Rowe in response to IE Bulletin No. 80-04 (Ref. 9-9).
IE Bulletin No. 80-04 requested a review of the main steam line rupture analyses support-ing plant operation to detennine if the assumptions made in the analyses regarding feedwater system operation were appropriate.
Four concerns were specified in the IE Bulletin No. 80-04 and are listed below:
- 1) Containment Pressure response.
- 2) Feedwater system pump (main, condensate, auxiliary) operability
- 3) Ability to detect and isolate a damaged steam generator, and
- 4) The potential for core return to power As a result of the Yankee review, which is documented in Reference 9-10, it was suggested to implement two design changes; 1) auto tripping of the con-densate pump on coincident high containment pressure and low steam line pressure,,
and 2) ensuring boiler feed pump auto trip on coincident high containment pressure i and low steam line pressure at power levels greater than 15 MWe,
Yankee Rowe believes these design changes are prudent since the changes would lessen the severity of a main steam line, rupture transient. Addition-ally, emergency procedures were modified to provide additional assurance of feedwater tennination to a damaged steam generator.
The findings of the above Yankee Rowe review have been included in the Cycle XV reload perfomance analyses. The licensee stated that the only differences between Cycle XIV and XV are core physics-related parameters.
Thus the first three issues cited above are bounded by the licensee's re-sponse to IE in Reference 9-10.
For Cycle XV, a review of the core physics-related parameters including moderator reactivity, Doppler re-activity and inverse baron worth was performed.
From this review, the licensee concluded that rod insertion limitations of the Technical Spe-cifications ensure adequate shutdown margin for all operations at critical l
conditions. The shutdown margin, consistent with rod' restriction limits, will preclude a post scram return to criticality. The licensee has con-cluded that operation within the proposed rod insertion restrictions and shutdown margin requirements will ensure that the consequences of this event remain acceptable for Cycle XV operation.
We have reviewed the licensee's evaluation reports with regard to the main steam line rupture accident (Reference 9-5, 9-6, and 9-10) and concluded that the evaluation is acceptable. However, the licensee has agreed to provide a confirmatory analysis of the main steam line rupture accident assuming concurrent loss of offsite power with Cycle XV physics parameters and system modifications to demonstrate that reactor system pressure, fuel performance and radioactive doses are within acceptable limits during the accident.
7.
Steam Generator Tube Rupture Event The steam generator tube rupture event is a failure of the recctor coolant system pressure boundary.
The reactor protection system pre-vents core damage.by actuating a reactor trip on low main coolant system pressure signals on two out of these channels.
In addition, a safety injection system is provided to replenish the coolant inventory lost through the ruptured tuLc.
The licensee has provided, in Reference 9-5, the results of an analysis which demonstrated that the reactor coolant systam pressure, fuel per-formance, and radioactive release after the postulated steam generator tube rupture event are within acceptable limits.
The analysis of this event for the reference Cycle XI also demonstrated that the results are not sensitive to the core design. Therefore, the results of the reference analysis will also be applicable to the Cycle XV reload. He find this acceptable, 8.
Loss-of-Coolant Accident (LOCA) a.
Small Break LOCA Core XV operations are conservatively bounded by previous small break analyses for Core XIII.
The Core XIII analyses are based on a peak linear heat generation rate of 12.85 kw/ft.
The Core XV allowable peak linear heat generation rate is limited-to 11.5 kw/ft, due -to large break LOCA results.
b.
Large Break LOCA Core XV large break LOCA calculations were performed using YAEC's UREM-based ECCS Evaluation Model. The following changes were made to
~
~
the model: (1) the cladding swelling and rupture model data was obtained from NUREG-0630 (Reference 9-1). (2) the ENC-WREM-II (Reference 9-2) flow rate multiplier curve for reduded flow area was used to be consistent with the NUREG-0630 local blockage curve, and (3) a lower plenum phase separation model (Reference 9-3) was incorporated into the blowdown calculation. The lower plenum model has been approved by the staff (Referenc
- 9-4).
The peak cladding temperature occurs during the reflood portion of the transient.
YAEC was contacted by the staff to confirm that the analyses provided correctly accounted for any plugged steam generator tubes. The analyses prov ded does account for the current number of plugged steam generator tubes.
The analysis presented by the applicant for Core XV operation, within the limits specified by the analysis, have been shown to yield LOCA results that are within the limits specified in 10 CFR 50.46.
9.
References 9-1.
D. A. Powers and R. O. Meyer, " Cladding Swelling and Rupture Models for LOCA Analysis," NUREG-C530, April 1980.
9-2.
XN-76-27, "lxxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC-UREM-II," July 1970.
9-3.
G. J. Brown, et al, " Application of a lower Plenum Phase Separation Model to Yankee Rowe Large Break LOCA Ar.ilysis," YAEC-1231. Rev.1, November 1981.
o n.
i 9-4.
Memorandum for G. C. Lainas, A/D for Safety Assessment, DL.
from P. S. Check, A/D for Plant Systems, DSI, Yankee Rowe Reload Cycle XV," June 9,1981.
9-5.
Proposed Change No.115, " Core XI Refueling," submitted DOL /AEC on March 29,~1974.
9-6.
Yankee Nuclear Power Station Core XV Performance Analyse.e.
thrch, 1981.
9-7.
Letter, WYR 78-99, dated November 21, 1978, D. E. Vandenburgh to USNRC, " Additional Information - Core XIV Refueling."
9-8.
Letter, WYR 79-104, dated September 12, 1979, D. E. Vardenburgh to USNRC," Yankee Rowe Loss of Feeodater Analysis."
9-9.
Letter, ilSNRC To YAEC. dated February 8,1980.
9-10.
Letter, WYR 80-50 D. E. Moody of YAE to USNRC, dated May 8,1980,
" Response to IE Bulletin No. 80-04."
. C.
RjACTORFUELDESIGN 1.
Introduction By letter dated March 26,1981 (Ref.1), the Yankee Atomic Electric Company (YAEC) made application to modify the Technical Specifications for Yankee Rowe to permit operation for.a fifteenth cycle. The Cycle 15 reloed appli-cation involves a fuel design similar to that previously considered nr Yankee Rowe.
In addition to a number of changes in the fuel system design, the application also contains a number of changes to the methods used in the plant safety analysis, as reported in Reference 2.
Our evaluation of this application follows.
2.
Fuel Systems Design The Yankee Rowe Cycle 15 core will consist of 76 fuel assemblies with fuel rods arranged in 16x16 arrays. Each fuel rod is composed of a number of U0 fuel pellets in an eight-foot-long Zircaloy-4 tub.e. The outer row of 2
fuel rods in each array are only partially filled with rods to allow room for insertion of cruciform control roos between assemblies. The structural integrity of each assembly is maintained by six spacer grids and eight guide bors. The metal guide bars, or tic rods, replace fuel rods in the outside row of each assembly and provide structural support in the vertical direction. The assembly is not shrouded (Ref. 3), as was the case for several earlier fuel designs used at Yankee Rowe.
The ',ankee Rowe Final Safety Analvsis Report (Ref. 4) continues to show the older design. A discussion of the current design, including revised fuel design drawings, can be found in Reference 5.
i
The 36 fresh fuel assemblies loaded for Cycle 15 operation differ in several respects from the 40 previously irradiated assemblies reloaded for this cycle. The spacer grids, grid straps, guita bars, instrumentation j
tubes.and thimbles are now fabricated from Zircaloy-4 rather than 304 stainless steel. This material change reduces parasitic absorption of i
235 neutrons and allows the use of slightly lower U content (3.57. vs. 4.07,)
in the fuel.
In order to allow the use of Zircaloy, rather than stainless steel, grid straps without reducing the mechanical strength of the design, the grid j
strap thickness was increased from 17 to 20 mils. Spacer grid design and I
grid crush strength tests were perfonned by the fuel supplier, Exxon Nuclear, to verify (Ref 6) that the structural strength of the Zircaloy grid is at least that of the previous stainless steel grid. The overall fuel assembly length has been increased slightly (111.775 inches vs. 111.55 Zircaloy. An improvement in the fuel thennal perfornance characteristics was i
also obtained by increasing the fill gas pressure from 143 to 250 psig; i
Additional, mi.nor modifications to the spacer grid, tie plate and holddown spring system have been incorporated (Ref. 7). We have reviewed these fuel design changes and have concluded that an appropriate analysis of these changes has been perfonned. We therefore find the Yankee Rowe Cycle 15 fuel j
system design acceptable.
l 3.
Fuel Thermal Design As discussed in Section 2 of this report, the fresh fuel in the Yankee Rowe Cycle 15 core is similar to that previously irradiated in the reactor. The L.
e licensee's analysis of the fuel thermal performance is also the same as that used in previous reload analyses with two exceptions:
(1)theanalysis now considers a number of' power history effects and (2) the analysis now considers burnup-dependent fission gas release as prescribed in NUREG-0418 (Ref.8).
The powr history effects relate only to the calculation of rod internal pressure and fuel centerline melt limits.
In the past, the licensee con-sidered a lead rod in which the power rating and bornup bounded the expected values of these parameters for the core in question.
In the revised analysis, the licensee has recognized the fact that maximum power and maximum burnup do not, in practice, occur in the same rod when previously irradiated fuel is present. To consider this feature in the fuel thernal analysis, it is necessary to consider a number of power histories. Each power history is limiting for power or burnup (not both) for each fuel type in Cycle 15 core.
Because it is no longer obvious which history will produce maximum fuel temperatures or rod internal pressures, all results must be examined to find the maximum conditions. The licensee has performed such an analysis for Cycle 15 operation, considering previous cycle exposures and uncertain-ties. The results show that (1) calculated internal fuel rod pressures are less than minimum operating coolant system pressure throughout Cycle 15 operation and (2) beginning-of-life conditions continue to yield maximum predicted fuel temperatures.
The licensee's use of a burnup-dependent fission gas release model is the result of an NRC request (Ref. 9) to all U.S. fuel vendors to consider this l
e c
_ ig.
effect.
Because the licensee has elected to use a method provided by the staff (Ref. 8) to consider this effect, we consider further review of this change unnecessary.
We, therefore, conclude that the fuel thermal design analysis for Yankee Rowe Cycle 15 is acceptable.
4.
Thermal-llydraulic Design A single fuel-related issue is discussed here; this issue concerns the rod bow magnitude, which was not contained in the licensee's description of the Cycle 15 thermal-hydraulic design (Sect;on 6 of Ref. 2). The licensee has cited References 10 and 11 as a basis for rod bowing analys,is in the Cycle 15 reload application. Reference 10 is a transmittal from the fRC staff in which the departure from nucleate boiling ratio (Df1BR) reduction due to rod bowing at Yankee Rowe is assumed to be con;enstated for by other thennal-hydraulic margins. These margins are identified in Reference 11.
Although the actual margin necessary to cover any DNBR reduction was not identifed in either reference, the staff's interim safety evaluation on rod bow (Ref.12) applied a full closure (rod-to-rod contact) DNBR reducticq of 347. based on Westinghouse data. The DNBR margins identifed in Reference 11 are greater than the DNBR reduction based on rod-to-rod contact. Thus, no penalty was required for the three previous cycles of operation at Yankee Rowe.
In the absence-of a specific relationship for rod bow magni-tude versus burnup for fuel, we continue to find the full closure DNBR penalty appropriate for Cycle 15 operation. The continued applicability of the thennal hydraulic margins identified in Reference 11, and the manner in which
e.
1 i
.they are used to offset the rod bow penalty, are discussed in Section D of this. Safety Evaluation Report.
5.
Operating Experience 5.1 Fuel Failures As a result of high coolant activity levels measured during Cycle 14 operation 4
at Yankee Rowe, some fuel rod failures were' suspected, and an examination was initiated by the licensee (Ref.13) during the reload outage. The licensee has infonaed us that all fuel assemblies but one were sipped. The exception was assembly B574, which had a visible indication of fuel failures. The sipping results gave an indication of fuel failure in only one additional assembly, B550.
Assembly B550 is a second-cycle assembly that would normally have been discharged at this time. This assembly was discharged as planned.
The visually examined assembly, B574, indicated several missing fuel rod end plugs (with springs protruding) and four-pins sheared and bent (circumferential breaks). This assembly was scheduled for reinsertion in Cycle 15.
All rods were removed from the assembly and eddy-current tested. The eddy-current testing indicated five additional rods with cladding wear (fretting). The sound rods were placed in a new fuel assenbly cage with Zircaloy-clad, Zircaloy-filled dummy rods replacing the nine failed or suspect rods. The reconstituted assembly will be reinserted for Cycle 15 operation.
The licensee attributes (Ref.13) the damage to vibration-induced fuel rod fretting. The damage was confined to the upper portion of the fuel assembly and appears to be related to previous fuel failures (Ref.14). Both the
~-
. current and previously reported failures appear to be limited to fuel assenblies positioned next to the core baffle (see Figure 1). As a
. result of the fretting problem, the licensee has replaced fuel rods in several suspect locations with dummy rods. These dummy, or sacrificial, rods are Zircaloy-clad stainless steel pins in the outer row of fresh fuel assemblies rer.iding on the WEST SIDE of the core p:riphery (see Figure 2). These fresh assemblies also have a slightly modified grid design (see Section 2). Two fue? rods will be replaced in Type A assemblies at core locations A4, B3, C2 and 01. Four fuel rods will be replaced in Type B assemblies at core locations A7, B8, C9, and 010.
Should cladding failure occur in one of these non-fueled dumny rods, no increase in coolant activity levels would be expected. Furthermore, the peripheral location of the metal dummy rods makes them, from the physics standpoint, nearly indistinguishable frot the core baffle.
Although we agree with the licensee on the method used to reduce the number of new failures, there is no assurance that all susceptable rod Iccations are now filled with duany rods. Furthermore, the actual damage process has not been identified and may not be affected by the modified grid design. We conclude tha* potential fuel failures during Cycle 15 operation, although small in number, may not be entirely eliminated and that continued surveillance of this problem is required.
In response to our concern, the licensee has committed (Ref.15) to continue monitoring coolant activity levels during Cycle 15 operation and to notify the staff of significant changes which may indicate additional failures. Asin I
. the past, the licensee will continue an investigation into the cause and prevention of such failures, and make further repairs to damaged fuel as necessary.
In view of the small number of failures observed to date, and the licensee's continued efforts to resolve this problem, we find the issue of fuel failures during Cycle 15 operation adequately addressed.
- 5. 2 Control Rod Bow During a controlled plant shutdown for refueling at the end of Cycle 14, control rod number 17 became inoperable and would not scram upon loss of stationary gripper coil voltage (Ref.16). The remedial actions required by the plant Technical Specifications (driving the rod into the core by operating the drive mechanism in reverse) were followed and a large shut-down margin was maintained. The cause of the problem was subsequently identified as a bowed Zircaloy follower on the control rod.
The twenty-four control rods enployed in Yankee Rowe are cruciform shaped and are inserted between fuel assemblies (see Figure 2). Two of the control rods are bare hafnium metal with stainless steel rubbing plates. The remaining control rods are of a newer design using a silver 'ndium-cadmium alloy with an Inconel Jacket.
Both designs employ a Zircaloy extension, or follower, which is also cruciform-shaped. When the control rod is pulled (up) out of the : ore, the Zircaloy follower is pulled into the core to avoid a large " water hole" of moderating coolant. When a control rod,is inserted, the follower moves down out of the core into a control rod extension tube in the lower plenum (see Figure 1).
4 r,
cf.
During nonnal operation, the neutron absorbing section of the control rod is withdrawn and the Zircaloy follower is resident in the active core.
Zircaloy is known to grow (physically deform) under irradiation, and the Zircaloy followers used at Yankee Rowe exhibit this' effect after a period of time. The follower of control rod number 17 became sufficiently bowed from the irradiation growth to stick when insertion was attempted. The licensee has informed us that the rod was near the full insertior limit at the time it stuck and the point of hang-up probably at the lower core plate. The clearances involved are on the order of 0.4 inches.
In correcting this problem, the licensee has noted that the followers used 4
on the older hafnium rods have a different metallurgical history than on the more recent silver-Indium-cadmium design. The older design was annealed and straightened during fabrication, whereas the ne.ser design was annealed, straightened, and stress-relieved. The newer process is less susceptible to the irradiation-induced growth phenomenon. As a result of this infor-mation, the licensee has replaced b th hafnium control rods with fresh 9
rods of the newer design.
The licensee has also performed a 100% inspection of the modified core com-planent of silver-indium-cadnium rods. This inspection involved visual measurement of control rod bow with a transit as well as measurement of control rod drop times and withdrawal forces. The visual measurement of control rod bow is a difficult task, with measurement uncertainties of approximately
.o
+ 0.1 inch. Based on an acceptance criterion of 0.2 inch or less, seven additional control rods were replaced with new rods.
The measured control rod drop times did not show a strong correlation with measured bow. All drop times were shown to be within acceptable limits. The rod withdrawal forces, however, do show a significant correlation with measured bow as shown in Figure 3.
Based on these data, three additional rods (for a total of twelve control rods) _were I
replaced with new control rods.' We further note that, during each o
refueling outage, the control rods are rc sted 90 (in position) to t
obtain a more uniform irradiation history. Approximately 1/4 of the 3
rods in the core are subject to visual surveillance during each outage.
This surveillance program, in place since 1972, was apparently not sufficient to avoid the recent stuck rod occurence. Our review of simil'ar events at Yankee Rowe shows that a previous stuck rod (1979, Ref.17) was caused by crud buildup in the control rod drive mechanism rather than rod bowing. Based on the improved follower design now used,
[
and the licensee's extensive surveillance during the current outage, we conclude that probability of a stuck rod during Cycle 15 operation has been signficantly reduced. We find the licensee's actions to date adequate and will address the issue of additional surveillance-in our review of the Cycle 16 submittal.
6.
Transient and Accident Analysis
~ The analysis of transient and accident conditions for Yankee Rowe Cycle 15 generally follows methods and analyses previously approved b5 the NRC. There i
- - ~
are, however, several revisions in the methodology use which necessitated I
a reanalysis of certain events. The' revisions related to the fuel system are discussed below.
6.1
. Loss-of-Coolant Accident Three fuel-related items were ~ addressed by the licensee for the postulated loss-of-coolant accident (LOCA). These were (1) Cycle 14 versus Cycle 15 design differences, (2) cladding swelling and rupture as described in flVREG-0630 (Ref.18), and (3) enhanced fission gas release as described in tiUREG-0418 (Ref. 8).
To evaluate the impact of these changes on the LOCA analyses, the licensee perfonned both a break spectrum arialysis and a burnup sensitivity study.
A partial, rather than complete, LOCA re-analysis was performed because I'
enhanced fission gas release effects do not occur for burnups below 20 GWD/MTU. The results show that the Yankee Rowe Cycle 15 core continues to satisfy regulatory requirsaents for LOCA analysis.
7.
Conclusior.s We have reviewed the fuel system design and analysis for Yankee Rowe Cycle 15 operation and find the application acceptable.
= _ - _
i 26 -
1 REFEREfiCES 1.
L. H. Heider (YAEC) letter to the Office of Nuclear Reactor Regulation, (NRC) on " Core XV Refueling" dated March 26, 1981.
2.
J. R. Chapman et al. " Yankee Nuclear Power Station Core XV Performance Analysi s," Yankee Atomic Electric Company Report YAEC-1240, March 1981 (Attachment D to Reference 1 above).
3.
J. L. French (YAEC) letter to the Office of Nuclear Reactor Regulation, (NRC) dated July 15, 1975.
4.
Yankee Nuclear Power Station Final Safety Analysis Report, Yankee Atomic Electric Company Report dated January 3,1974, U. S. Nuclear Regulatory Commission Docket Number 50-29.
5.
D. E. Vandenberg (YAEC) letter to the Office of Nuclear Reactor Regulation, (NRC) on " Core XII Analysis" dated November 7,1975.
l 6.
T. J. Helbling (Exxon) letter to J. Kay -(YAEC) dated June 25, 1981.
7.
D. A. Adkisson (Exxon) letter to R. T. Chin (YAEC) dated February 7,1979.
8.
R. O. Meyer, C. E. Beyer and J. C. Voglewede, " Fission Gas Release From Fuel at High Burnup,"
U. S. Nuclear Regtlatory Conmission Report NUREG-0418, March 1978.
9.
D. F. Ross (NRC) letter to W. S. Nechodom (Exxon) dated January 18, 1978.
10.
A. Schwencer (NRC) letter to R. H. Groce (YAEC) incorrectly dated January 4, 1976 (issued January 4,1977).
11.
D. E. Edwards (YAEC) letter to the Office of Nuclear Reactor Regulation (NRC) on " Fuel Rod Bow Effects on Margin to DNB at Yankee Nuclear Power Station" dated February 9,1977.
. 12.
D. F. Ross and D. G. Eisenhut (HRC) memorandum for D. C. Vassallo and K. R. Goller (NRC) on " Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on Thennal Margin Calculations for Light Water Reactors" dated December 8,1976 (Attachment to Reference 10 above).
13.
H. A. Autio (YAEC) letter to B. H. Grier (NRC/ Region I) transmitting Licensing Event Report 50-29.81-03/0lT and dated June 10, 1981.
14.
H. A. Autio (YAEC) letter to B. H. Grier (NRC/ Region I) transmitting Licensing Event Raport 50-29/78-31/0lT ard dated November 25, 1978.
15.
J. A. Kay (YAEC) letter to D. M. Crutchfield dated July 15, 1981.
16.
H. A. Autio (YAEC) Letter to B. H. Grier (NRC/ Region I) transmitting Licensee Event Report 50-29/81-05/03L and dated June 1,1981.
17.
H. A. Autio (YAEC) letter to B. H. Grier (NRC/ Region 1) transmitting Licensee Event Report 50-29/79-02/03L and dated February 21, 1979.
18.
D. A. Powars and R. O. Meyer, " Cladding Swelling and Rupture Models for LOCA Analysis, U. S. Nuclear Regulatory Cocaission Report NUREG-0630, April 1980.
l
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28 -
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, D.
THERMAL HYDRAULIC ANALYSIS' l.0 Thermal-Hydraulic Design The Yankee Row Cycle XV core containe % fresh assemblies and 40 recycle assemblies. As in Core XIV, Core XV is entirely composed of Exxon fuel.
The core average Nrnup for the begining-of-life Core XV is 6,917 MWD /MTU compared to 5,268 PWD/MTU for Core XIV. The full power lifetime of Core XV is 11,500 MWD /MTU compared to 14,100 MWD /MTU for Core XIV.
The fresh assembly grid spacers are 3 mils thicker than the grid spacers ir the recycled fuel.
The objective of the review is to confirm that the thermohydraulic design of the reload core has been accomplished using acceptable methods, and provides acceptable margin of safety from conditions which would lead to fuel damage during nomal operation and anticipated operational transients.
- 2. 0 Evaluati6n The thermal-hydraulic evaluation of the reload cycle XV has been performed utilizing COBRA 3C (Ref. 3) and W-3 critical heat flux correlation (Ref. 4) as in the previous four reload analyses (Cores XI, XII, XIII, and XIV) with the following exceptions:
1.
Fuel thermal performance analysis, including the calculation of pellet and clad temperatures and the consideration of cladding integrity and collapse, has been performed as described in Section 4.0 of Reference 2.
t 2.
Predicted hot channel factors are based on t'eginning-of-life (80L) power distributions obtained when Rod Group C is 25 percent inserted even though rod restrictions do not permit operation at full powr l
in this mode.
\\
- Core XI serve: as the reference thermal-hydraulic analysis, as hc5 been the case for Cores XII through XIV, since a canplete safety analysis was performed for Core XI. The thermal-hydraulic analysis of the reload cycle was performed by adjesting code input to reflect the reload cycle power distributions and thermal-hydraulic characteristics.
A canparison of thermal hydraulic design conditions for Yankee Rowe Cycle XV and cycle XI for 4-loop operation and 3-loop operation is provided in Tables 1 and 2 (obtained from Tables 6-1 thru 6-6 in Ref. 2). The reload core has significant margin to DNB, coolant quality and fuel centerline melt limits. The design DNBR.for the recycled fuel is essentially the same as the Core XI Reference DNBR, 3.23 versus 3.24 at full power 4-loop operation, and 3.97 versus 4.00 during 3-loop operation.
The Design DNBR for the fresh fuel is marginally less than that of the recycled fuel and Core XI fuel, 3.13 versus 3.24 at full pover 4-loop operation and 3.86 versus 4.00 during 3-loop operation.
This small difference, 3 percent, occurs because of small differences in flow characteristics between fresh and recycled fuel assemblies. The fresh assembly grid spacers are 3 mils thicker than the grid spacers in the recycled fuel. The impact of this increase in grid spacer thickness is an overall increase in assembly flow resistance.. Fresh assembly flow rates are marginally less than recycled assembly flow rates and the impact of this reduction is marginally lower DNBR's and slightly hight: hot channel exit tenperatures for design condi-ti ons.
Note, however, that predicted performance of both fresh and recycled w
w-1 w
33 -
fuel are essentially identical and within Core XI predicted thermal perfor-mance. - This situation exists because fuel power distributions are more i
favorable than recycled fuel power distributions.
2.1 R6d B6w Pdnalty The staff interim safety evaluation on rod bow (Ref. 7) applied a full closure (rod-to-rod contact) DNBR reduction of 34% based on Westinghouse data. Margins available to DNB (Ref. 5, 6) for Cycle XV core are given in Table-3. These margins are greater than the 34% reduction required due to rod bow.
Reference B confirms the applicability of these' margins to Yankee Rowe Cycle XV operation. These margins have been reviewed by the staff and were found to be acceptable. Therefore, no rod bow penalty is required for Cycle XV.
- 3. 0 Evaldatidn'6f the Transients The licensee has rqviewed the anticipated occurrences and the postulated accidents whica were reported in Reference 9.
Each transient was considered and compared with the analysis presented in the above reference which was previously approved by the NRC (Ref.10). For those incidents which were rot bounded by the previously approved analysis, new safety analyses were provided which demanstrated that the applicable design basis limits were not. exceeded.
Table 4 provides the initial operating conditions for most of the transients.
Minor differences betwaen reload core and reference cycle exist in basic plant parameters. These minor differences are the following:
q
---e
, o 1.
Core XV Core Inlet Tenperatures is 4 F higher than the reference cycle o
o o
(515 F versus 511 F).
Plant operation at 515 F core inlet tennpernure was approud during the Core XIII reload submittal and has occurred during both Core XIII and Core XIV.
The reviews perfonned during the Core XIII o
and XIV reloads demonstrated the minor impact of the 4 F increase (Reference 11 and 12).
2.
Maximum linear heat rate and hot channel factors are reduced from Core XI values. Core XV.:alues are identical to Cores XIII and XIV.
In each case, Core XV values are more favorable than Core XI reference cycle values 3.
Minimum DNB ratios at design conditions for Core XV fresh fuel are marginally less than the Core XI reference analysis value. The reasons for this difference were discussed in Section 2.0.
The staff has review 6d the initial conditions, given in Table-4. used in the transients. The limiting transient for Yankee Rowe with respect to DNB is the loss of flow in 2 out of 4 pumps (Reference 6).
This Eeent re-suits in a minimum DNBR in excess of 2.05, which is larger than the 1.64 value needed to,accommo,date the rod bowing penalty, and this is acceptable to the staff.
4.0 Technical Specifications 4.1 Tha licensee has proposed modifications to Reactor Core Safety Limit curves, figures 2.1-1 and 2.1-2 in attachment B. Reference 1.
These curves bound both fresh and recycled fuel assemblies. The licensee modified these curves conservatively in an effort to bound future core characteristice and,hence, L
w
< eliminate future technical-specification changes in this area. These pro-posed-modifications to the Safety. limit curves have been' reviewed by the staff and are found.to be acceptable..
- 5. 0 Conclusion We have reviewed the thernal hydraulic design of Core XV and-conclude that in most cases the difference in thermal hydraulic parameters between Core
- l XI and Core XV is insignificant.. We conclude that the thermal hydraulic design of; Core XV meets the design criteria and is acceptable.
i i
\\
l l
i 4
O I
I i
. TABLE-1 Thermal Hydrddlid' Papsmeters for Yankee Rowe Cydid XV 's Cydie 'XI* 'Ddring 4-Loop 6peration General Characteristics Predicted Design Total Core Power, MWt 600 618 (600 )
(618)
. Main Coolant Pressure, Psig 2000 1925
-(2000)
(1925) o-Main Coolant Inlet Temperature, F 515 519 (511)
(515)
Total Coolant Flow Rate, lb/hr 38.3x106 38.3x10 6 6
6 (38.3x10 )
(36.3x10 )
Nominal Channel Hydraulic Diameter, in 0.412 0.412 (O.399)
(O.399) 2 0
Average Mass Velocity,1b/hr-ft 2.29x10 2.29x10 6
6 (2.36x10 )
(2.36x10 )
2 Average Heat Transfer Area, ft 167 167 (171)
(171)
- Values in parentheses i
{
o TABLE cont.
Hot Channsi and Hot' Spot Parameters Prsdicted Design Fresh Recycled Fresh Recycled Fudl
' Fusi Fuel FdeT Maximum Centerline Pellet 2955 2484 3419-2756 OF (3430)
(3770)
Temperature Minimum W-3 DNB Ratio 4.52 A. 53 3.13 3.23 (4.48 )
(3.24 )
Total Heat Flux Factor 2.38 2.41 2.76 2.76 (2. 67 )
(2. 96)
Total Enthalpy Rise Factor 1.47 1.49 1.82 1.78 (1.76 )
(1.81) i 4
m
- r'l
. 38 -
TABLE 2
. Thomal LHydraulic Parameters. for Yankee Rowe Cycle XV vs Cycle XI* During 3-Loop Operation General Chdrdcteristics Prsdictsd Dssign Total Core Power, MWt 450 463.5~
(450 )
(463.5)-
Main Coolant Pressure, Psig 2000 1925 (2000)
(1925) o Main Coolant Inlet Temperature, F 515 519 (511)
(515)
Total Coolant Flow' Rate, lb/hr 29.9x10 (29.9x10 )
2 Average Mass Velocity, lb/hr-ft 1.79x106 6
(1.84x10 )
Hot Chdrinel'and Hot Spot Pararnetsis Prsdicted Design l
Fresn Recycled Fresh Recycled' Fuel Fuel Fuel Fusi Maximum Centerline o
- 1 Pellet Temperature, F 2339 2041 2706 2297 (2770)
(3080)
Minimum W-3 DNB Ratio 5.40 5.41 3.86-3.97 (5.35)
(4. 00) l
' Total Heat Flux Factor 2.38 2.48 2.76 2.76 (2. 67 )
(2. 96 i Total Enthalph Rise Factor 1.47 1.48 1.80
.1. 76 -
(1. 78 )
(1.81)
Values in parenthesis a
[
l l.
TABLE 3 Credits Applicable to Yankee Rowe to of fset D:.~3R Penalty Resulting from Fuel Rod Bow Percent D!GR Parameter Desien Value Calculavad Value Marcin t!
F 1.80 1.61 (0-10,000 :n D/MTU) 4.5 3g 1.61-1.'44 (10,000-30,000 4.5-9.6 E*D/MTJ)
F 1.80 1.73 (0-10,500 WD/MTU) 4.0 XY 1.73 -1.49 (10,500-30,000 4.0-20.8 s'D/MTU)
F^
~
1.02 1.0 2.0
- 8 1.07 1.0 2.7 F
AH Minict:n CI:ER resulting from lir.iting D'!2R transient 2.05 36.6 N
Design nuclear enthalphy rise factor F
3H Design one pin peak power F
xy AUG F
One pin peak power Q
0 F
Hot channel enthalpy rise
1
- 40~-
TABLE 4 INITIAL OPERATING CONDITIONS Reference Cycle Reload Cycle Core II Parameter (4-Loop /3-Loop)
(4-Loop /3-Loop)
Reactor Power, MWt 600/450 600/450 Core Inlet Temperature, F 515/515 511/511 Main Coolant Pressure, psia 2015/2015 2015/2015 Minimum Reactor Coolant Flow, 35.0/27.3 35.0/27.3 106 lb/hr Maximum Linear Heat Rate, 12.5/9.38 12.9/9.68 kw/ft Axial Heat Flux Profile Cosine / Cosine Cosine / Cosine Total Heat Flux Factor 2.76/2.76 2.96/2.96 Total Enthalpy Rise 1.82FF, 1.78RF/1.80FF, 1.76RF 1.81/1.01 Factor Maximum Pellet centerline 3419/2706
>3419/>2706 Tempe ra ture, F Minimum W-3 DNB Ratio 3.13FF*,3.23RF*/3.86FF,3.97RF 3.24/4.00
- FF = F'resh Fuel, RF = Recycled Fuel 1
e 4
0
l f REFERENCES l
1.
Yankee Atomic Electric Company letter FW181-52 (L. H. Heider) to NRC
" Core XV Refueling," dated March 26, 1981.
2.
YAEC-1240, " Yankee Nuclear Power Station Core XV Performance Analysis,"
March 1981.
3.
A. E. Ladieu, "A Thermal-Hydraulic Analytical Model Using COBRA-3C" YAEC-1058 (May 1974).
4.
L. S. Tong, "DNB Prediction for an Axially Non-Uniform Heat Flux Distri-bution," WCAP-5584, (1965).
5.
Letter A. Schwencer (NRC) to R. H. Groce (YAEC), " Yankee-Rove Atomic Power Station, dated January 4,1976.
6.
Letter WYR77-13, D. W. Edwards (YAEC) to NRC, Fuel Rod Bow Effects on Margin to DNB at Yankee Nuclear Power Station, dated February 9,1977.
7.
D. F. Ross and D. G. Eisenhut (NRC) memorandum for D. D. Vassallo and K. R. Goller (NRC) on " Interim, Safety Evaluation Report cn the Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors" dated DecenJuer 8,1976.
8.
Letter, R. Caruso (NRC) to Branch File, " Core XV Reload, dated July 9,1981.
9.
Proposed change No.115, " Core XI Refueling", Submitted to DOL /AEC on March 29,1974.
10.
Amendment No. 9, letter from K. R. Goller, DOL /AEC to YAEC, Attn: G. C. Andoymini, July 30,1974.
- 11. Proposed change No.145, Supplement 1, Submitted to USNRC April 13, 1977.
- 12. Proposed changes No.163, Submitted to USNRC September 8,1978.
1 1
I r
e
e 9 '
E.
Main Steam Non Return Valves - Mechanical and Transf er.t Analysis -
~ The four main steam line non-return valves (one associated with each steam generator) provide a check function against back flow of stear, from the turbine, but must be closed mechanically to prevent forward flow.
Prior to the Cycle XV reload, mechanical closure has been performed manually.
For' Cycle XV, and subsequent cycles, automatic closure equipment has been installed.
In the previous arrangement, a steam line break downstream of the non-return valves would result in blow down of all four steam generators. The present arrangement with automatic closure permits isolation of a broken steam line or steam jenerator from the intact ones, whether the break is inside or outside of containment.
The change was motivated by internal review by Yankee-Rowe subsquent to Three Mile Island in which they concluded that the temperature transient induced stress would be reduced under the automatic isolation procedure.
The licansee determined that atmospheric dump valves must be relocated from downst.eam to upstream of the automated non-return valves. Steam supply for the steam driven emergency feedwater pump must also be taken upstream of the automated non-return valves. These changes, including necessary isolation valves on the atmospheric dump lines and steam driven pump lines, have been instituted during the current shutdown.
No analyses of the revised system have been submitted, since the original FSAR steam line break analysis with all four stream generators blowing down is more limiting with respect to return to criticality.
- 43'--
The staff concludes that the steam system is upgraded by these changes.. No r.
I degradation of the return - to - criticality margin is anticipated, and' the thermal transient, pari'cularly.as reflected back into the primary system will certainly.
~
-be reduced. Availability of staam - driven feed pumps and atmospheric steam dump capability are not impaired.
The staff concludes that the modifications to the steam system non-return valves are acceptable.
G 1
f.
. :F. -AUXILIARY FEEDWATER SYSTEM MODIFICATIONS l '. 0 Introduction and Backgrcund
.The Three Mile Island Unit 2 (TMI-2) accident and :ubsequent investigatior.s and studies highlighted:the importance of the Auxiliary Feedwater System (AFWS) in. the mitigation of transients and accidents. As part of our-assessment of the TMI-2 accident and related implications for operating plants, we evaluated the AFW system for_ Yankee-Rowe and published the results as part of NUREG-0611. Our evaluation of this design is contained i-in the NUREG along with our recommendations-for the plant and 'the concerns i
which led to each recommendation. The objectivas of the evaluation were to: (1) identify necessary changes in AP4 system design or related procedures at the plant in order to assure continued safe operation, and j
(2) to identify other system characteristics of the AFW system which, on a-long term basis, may require system modifications. To accomplish these L
objectives, we:
-(1) Reviewed the AFW system design in light of current regulatory Tequire-ments (SRP) and, (2) Assessed the relative reliability of the AFW system under various loss 3
I of feedwater transients (one of which was the initia':ing event of TMI-2) and other postulated Tailure conditions by 6termining the potential for AFW system failure due to common causes, single point vulnerabilities, and human error.
We concluded that the implementation of the recommendations identified during this review could be expected to improve the reliability of the AFW system at Yankee-Rowe.
. 'Simularly,-the licensee determined that-additional changes were needed and made plans to install the modifications during the Core XV refueling outage. The purpose of this evaluation is to discuss the acceptability of these modificatior.s and their associated technical spec;i scations'.
2.0- Description of Changes These modifications increase the. existing emergency feedwater capabilities by adding two full capacity motor driven-pumps. Additional piping will be installed to allow feeding the-steam generators with either pump through the normal feedwater piping via the steam driven emergency feedwater pump discharge header,'or through the blowdown piping via the ~ alternate emergency feedwater header. System actuation will be capable of being initiated from the control room or locally by operator action.
The two motor driven pumps will be located in the Primary Auxiliary Building -
and will have the capability to take suction ' rom either the primary water storage tank, TK-39, or the demineralized water storage tank, TK-1 The pump suctions will be normally aligned to TK-39.
The new pumps will use existing 300 HP motors.
These motors will be powered from the 2400 V buses No. 2 and 3.
Remote flow indication to each steam generator will be availaole in the control room for emergency feedwater flow through the normal feedwater path,
Lccal flow indication will be availabla in the combined pump minimum recirculation flow piping.
t 4
L u.
sc l Technical Specification changes are necessary to -incorporate the new pumps, revise the surveillance requir:::nts, and revise the containment isolation
~
valve list _ due to the piping modification.
This design change increases the capability and the redundance.of the-emergency feedwater system, resulting in an increased system reliability.
The revised system will consist of three pumps, two motor driven and one steam turbine driven, capable of feeding the steam aenerators through two separate feed paths. Also, the capability will now exist to initiate f
emergency feedwater, via the two motor driven pumps, manually from the control room.
3.0 Evaluation We have evaluated the licensee's new AFW system against the recommendations of NUREG-0611 for Yankee-Rowe and have made the following determinations:
l l
a.
Recommendation GS The licensee should. k open single valves or t
multiple valves in series in the AFW system pump suction piping and j
lock open other single valves or multiple valves in series that could interrupt all ARJ flow. Monti.ly inspections should be performed to verify that these valves are locked and in the open position. These inspections should be proposed for incorporation into the surveill.ince -
requirements of the plant Technical Specifications.
i i
The licensee stated in a letter dated June 30, 1981 that the two motor driven pumps would share a common suction line from the primary water storage tank (PWST). The existing system design has been modified I
i such that the turbine driven pump ct. take suction from either the j
PWST or the demineralized water storage tank (DWST). All the valves i
4 w
F 47 -
in the suction line to each pump that could cause a loss of suction to the pump are norma open and are locked in that position. New Technical Specifications for the pumps require that at least once every 31 days, each pump will be run to verify that the proper ruction valves are open and that water can be pumped through the recirculation line.
We have further considered the revised design which utilizes two locked open valves in the single suction line to the PWST.
Inadvertent closure of either of the valves will result in loss of suction to both of the motor driven pumps. The pumps, however, would be started manually and the operator would monitor two separate indications for each pump; pump motor current and flow to the steam generators.
The pumps will be I
started in sequence; should a pump fail to start its pump motor current indication would alert the operator to this condition and should a pump fail to provide flow the flow indication to the steam generators will be available to alert the operator. Assuming a problem exists for the first motor driven pump, the operator has time to determine the extent of the problem before trying to start the second motor pump.
Further, the turbine driven pump which takes its suction from the DWST through separate piping is available should both motor driven pumps fail to start.
We therefore conclude that the licensee's response to
- l l
{
this recommendation is acceptable.
1
v,.
)
b.
Recommendation GS Emergency procedures for. transferring to. alter-nate sources of AFW supply should be available to the plant operators.
These procedures should include criteria to inform the operator when, and in what order, the transfer to a1 ternate water sources;should.
take place. The following cases should be covered by the procedures:
- The case in which the primary water supply is not~ initially available. The procedures-for this case should include any operator actions required to protect the AFW system pumps against self-damage before water flow is initiated; and,
- The case in which the primary water supply.is being depleted.
The procedure for this ctie should provide for transfer to the alternate water sources prior to draining of the primary
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water supply.
The licensee responded in a letter dated December 21, 1979, that YAEC-would revise its operating procedures to include the transfer of.
auxilicry feed suction to alternate sources. The p ocedure would specify when and in what order. We have reviewed the most recent operating procedures regarding auxiliary feed and have determined that this commitment has been met.
We conclude that the licensee's response to this recommendation is t
acceptable, j
Recommendation GS_5_ - The as-built plant should be capable of providing c.
l the required AFW flow for at least two hours from one AFW pump train, l
independent of any alternating current power source.
If manual AFW
, system initiation or flow control is required following a complete loss of alternating current power, emergency procedures should be established for manually initiating and controlling the system under these conditions. Since the water for cooling of the lube oil for the turbine-driven pump bearings may be dependent on alternating current power, design or procedural changes shall be made to eliminate this dependency as soon as practicable.
Until this is done, the emergency procedures should provide for an individual to be stationed at the turbine-driven pump in the event of the loss of all alternating current power to monitor pump bearing and/or lube oil temperatures.
If necessary, this operator would operate the turbine-driven pump in a manual on-off mode until alternating current power is r.estored. Adequate lighting powered by direct current power sources and communications at local stations should also be provided if manual initiation and control of the AFW system is needed.
(See Recommendation GL-3 for the longer-term resolution of this concern.)
In response to this recommendation, the licensee indicated in a letter dated December 21, 1979, that the turbine-driven pump AFW in the Yankee-Rowe station is capable of providing adequate flow for at least two hours following a complete loss of AC power.
The turbine-driven pump is manually started and controlled using the existing 6parating procedures.
The turbine driven AFW pump bearings do not require cooling water and pump sealing water is provided by gravity flow from the demineralized water storage tank.
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Emergency lignting and'the plant communseation. system are provided at the pump' and at the control valve station. Emergency lighting is -
battery _ powered and the communication systea can be powered feca the vital bus.
L
- We conclude that the provisions available in the existing AFW system at Yankee-Rowe meet the requirements outlined in this recommendation and are, therefore, acceptable, d.
Recommendation GS The licen<ee should confirm flow path availability of an AFW system flow train that has been out of service to perform periodic testing or maintenance as follows:
- Procedures should be implemented to require an operator to determine that the AFW system valves are properly aligned and a second operator to independently verify that tne valves are properly aligned.
- The licensee should propose Technical Specifications to assure that prior to plant startup following an extended cold shut-down, a flow test would c. performed to verify the normal flow path from the primary AFW system water source tc the steam generators. The flow test should be conducted with AFW system-valves in their normal alignment.
In a letter dated December 21,1979, the licensee indicated that the operating procedures presently in use at the Yankae-Rowe Station require operational testing following maintenance verification of valve line up following system testing.
Furthermore, the proposed Technical
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.. Specifications require. that prior to startup from cold shutdown all' three AFW flow paths be verified both by valve line up and by flow testing.
In addition, operating procedures requiring a second operator to verify it; dependently that the valves are properly aligned have been provided.
We have reviewed the licensee's response to this recommendation and cor.clude that is is acceptable.
e.
Recommendation - The AFW surveillance tests should require that the normally closed manually operated valves in the connection between the charging pumps / safety injection pumps and the AFWS be cycled each quarter.
The licensee responded in a letter dated December 21, 1979 indicating that the operating procedures have been changed to require a quarterly exercise of the normally closed valves in the connection between the charging pumps, safety injection pumps ard the AFWS.
We conclude that the licensee's response to this recommendation is acceptable, f.
Recommendation - The licensee should provide redundant level indications and low level alarms in the control room for the AFW system primary water supply to allow the operator to anticipate the need to make up water or transfer to an alternate water supply and prevent a low pump suction pressure condition from occurring.
The low level alarm setpoint should allow at least 20 minutes for operator action, assuming that the largest capacity AFW pump is operating.
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. The licensee responded in a letter dated December 21, 1979, indicating that the Yankee-Rowe station has one pneumatic level instrumentation loop for the primary water supply with indication in the control room.
In addition, float switches on the primary water supply provide. input signals to one high/ low level annunciator in the control room. An additional electronic level instrument loop with indication in the control room iq.'tiated by a pressure switch in the pneumatic level loop has been installed.
Each low level annunciator setpoint will permit the operator at least 20 minutes to -hange over to an alter-nate source of water.
We conclude that the installation of an additional electronic level instrument loop to the existing AFWS in Yankee-Rowe meets the require-ments outlined in this recommendation and is, therefore, acceptable, Recommendation (This recommendation has been revised from the original g.
recommendation in flVREG-0611) - The licensee should perform a 48-hour performance test on all AFW system pumps, if such a test or continuous period of operation has not been accomplished to date.
Following the 48-hour pump run, the pumps should be shutdown and cooled down and then restarted and run for one hour.
Test acceptance criteria should include demonstrating that the pumps remain within design limits with respect to bearing / bearing oil temperatures and vibration and that pump room ambient conditions (temperature, humidity) do not exceed environmental qualification limits for safety-related equipment in the room.
In a letter dated July'3,1980, the licensee submitted the results of an endurance test for the turbine driven pump. We.have reviewed the test data and conclude that it is acceptable.
By letter dated June 30, 1981, the licensee committed to perform a 48-hour endurance test for the new motor driven pumps. We consider the licensee response to be acceptable. Verification of the test results will be done by the Yankee-Rowe Resident Inspector,
- h. - Recommendation - Safety-grade indication of auxiliary feedwater flow to each steam generator shall be 'provided in the control room. The auxiliary feedwater flow instrument channels shall be powered from the emergency buses consistent with satisfying the emergency power diversity requirements for the auxiliary feedwater system set forth in Auxiliary Systems Granch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9.
The licensee responded in a letter dated December 2', 1979, indicating that control grade flow indication for the AFWS has been provided in the control room. By letter dated March 31, 1981, the licensee indicated that the flow indication instrument met all of the safety grade requirements listed in NUREG-0737, Item II.E.1.2.
We conclude that the licensee's response satisfies the " control grade" requirements specified in the NUREG-0578 position and clarifications and is, therefore, acceptable. The qualification of the instrumentation j
to meet safety grade requirements will be the subject of a separate review.
.l 1.
Recommendation - Licensees with plants which require local manual.
realignment of valves to conduct periodic tests on one AFW system train, and there.is only one remaining AFM train available for operation should propose Technical Specification to provide.that.a dedicated individual who is in communication with the control room be stationed at the manual valves.
Upon instruction from the control room, this operator would realign the valves in the AFW system train from the test mode to its operational alignment.
This recommendation is no longer applicable to Yankee-Rowe-because, with the addition of the two new motor driven AFW pumps, there are now three AFW trains available, so that testing of one train would still tvave two trains available for service.
J.
Recommendation GL At least one AFW system pump and its associated flow path and essential instrumentation should automatically _ initiate AFW system flow and be-capable of being operated independently of any AC power source for at least two hours. Conversion of DC power to AC power is acceptable.
The licensee responded in a lette'. dated December 21, 19 79, indicating that the turbine-driven AFW pump at the Yankee-Rowe station is a bladed turhine driven reciprocating positive displacement pump.
It was not designed for sudden starts. Normal starting procedure has the pump recirculation line open. Thus the pump is started and brought up to speed without load. The load is gradually applied by opening the flow control valves and slowly closing the recirculating valva. Prior to opening the steam admission valve the casing draas are opened, the steam admission valve cracked and the casing is warmed up.
After ti.
-e 55 -
casing is warmed,- the drain is closed and the admission valve is opened. enough to roll-off-the turbine.. Speed is' increased slowly i
_ until the operator observes :that the governor 'has. taken over. The:
admission valve is then fully opened. A dedicated AFW system 3
1 operator has been proveue'd for this purp'ose. ' The operator has approximately one hour's' time to perform his duty since.the steam -
generators' at Yankee Rowe have relatively long dry out times.
Automatic initiation of the turbine driven pump is not possible.
We are continuing our evaluation regarding the need for automatic-initiation of-the new motor-driven pumps and until our evaluation is ' complete, we will hold this item open. However, resolution is not required prior to plant startup from the Core XV refueling, and we will address this issue in a separate evaluation, k.
Recommendation - Initiation of AFW flow (including flow f r om the backup systems-charging /SI) to the steam generators requires several local manual operator actions outside the Control Room.
Even though there is a reasonable time period (up to one hour before the S/Gs will boil dry) for oferator action and a dedicated operator, the licensee should improve the reliability of initiating AFW flows by providing the capability to start the pumps and open the valves of the AFWs by operator action from the Control Room. Local manual operation capability should be retained as a backup to remote manual operation capability.
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- 56 In'a letter dated December 21,.1979, the licensee indicated that Yankee Rowe would provide two additional safety grade electric motor driven pumps in-the modified AFWS design.
These pumps can be started from the control room as well as locally.
Further, the valves involved are either locked open or can be opened from the contrcl room. We therefore,. conclude that the licensee's response to thir recommendation is acceptable.
1.
Recommendation - A pipe break in the Main Feedwater header upstream of the control valves could cause loss of all AFW flow to all steam generators since the AFW pump and the charging /SI pumps connect to this header. The licensee should evaluate he consequences of a pipe break in this section of the MFW header and 1) determine any system-design changes or emergency procedures necessary to detect and isolate the break and direct the required AFW flow to the steam generators before they boil dry or 2) describe how the plant car be b ought to a safe shutdown condition by use or other available systems ft llowing such a postulated event.
The licensee responded in a letter dated December 21, 1979, indicating
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that in the event of a pipe rupture in the main feedwater header upstream of the control valvts, the charging pumps or safety injection pumps can be used to provide feedwater to all four steam generators via the blowdown lines. The licensee stated that this feed mechanism was specifically installed to insure feed capability in the event that i
normal feedwater piping ruptured.
Current olant operating procedures are t
available for this mode of operation.
o._
F We conclude that the licensee's response adequately addresses this recommendation and is,-therefore, acceptable.
m.
Recommendation - The air operated trip valve in the auxiliary steam.
header ~ which supplies steam to the turbine driven AFW pump closes upon receipt of a containment isolation signal. The licensee should review the design basis for this circuit logic to determine whether all events that can generate a centainment isolation signal should.n fact, shc'!own the AFWS. As a result of this review, describe any design :hrnges or procedures changes that will be proposed to assure AFW system and containment isolation capability.
The licensee responded in a letter dated December 21,1979, that steam supply valve (TV-405) to the emergency boiler feedwater pump closes on a containment isolation signal (CIS). The licensee has reviewed the design basis for this trip logic and concluded that this air operated valve should be included in what the NRC terms
" essential" valves which should not be tripped on a containment
(
isolation signal.
In addition a dedicated nitrogen supply accumulator has been installed for the purpose of manually reopening this valve in the event of its inadvertent closure.
We conclude that tha licensee's response adequately addresses this recommendation and is, therefore, acceptable.
a 1 n.
Recommendation - The licensee should evaluate the need for the charging pumps and associated instruments and controls to be normally supplied by the emergency electrical buses since the charging pumps are backups to the one AFW pump.
In a letter dated December 21, 1979, the licensee indicated that the charging pumps, when used as backups to the one AFM pump, L39 be manually operated from the emergency bus. The licensee pointed out that the emertency diesel capacity is not sufficient to provide normal power to the charging pumps and simultaneously supply the existing ECCS require-ments. Further, the addition of two motor driven pumps eliminates the need for the charging pumps to act as backup AFW pumps.
We conclude that the licensee's response adequately addresses this recommendation and is, therefore, acceptable.
o.
Recommendation - The plant is within the scope of the Systematic Evalua-tion Program (SEP). The following additior 31 long-term concerns have been identified by SEi, and are applicable.
1.
The Yankee Rowe Nuclear Plant including the AFWS will be reevaluated during the SEP with regard to internally and externally generated missiles, pipe whip and jet impingement, quality and seismic design requirements, and earthquakes, tornadoes, and floods.
2.
The Yankee Rowe AFWS is not automatically initiated and the design does not have capability to automatically terminate feedwater flow to a depressurized steam generator and provide flow to the intact steam generator. This is accomplished by manual valve operation, I
g either from the control room or locally. The effect of this will be assessed in the main steam line break evaluation for this plant.
This item is still being evaluated and will be addressed separately.
Since the Yankee Rowe plant is within the scope of the Systematic Evaluation Program (SEP) the additional long term concerns stated above will be evaluated by SEP.
p.
Basis for AFWS Flow Requirements - In our letter dated November 9,1979, 3
I we requested the licensee to respond to Enclosure 2 of our letter, regarding the AFWS flow design basis. The licensee, in a letter dated September 12, 1979, provided a response to Enclosure 2.
We have reviewed the licensee's b
response and conclude that it is acceptable.
4.0 Conclusion 0ith the e).:eption of long term issues in items h, j, and o, which will be resolved at a later date, the licensee's responses to our recommenda-tions have been satisfactory. We therefore conclude that the rv "* <:a tio n s and additions to the AFW system and the Technical Specifications t.or.;es necessary to support these modi fications are acceptable.
! G.
ELECTRICAL AND CONTROLS EVALUATION 1A Introduction The licensee proposed modifications to autcmate the main steam line non-return valves (NRV), to add a main coolant systen (MCS) high pres-sure reactor trip, and to modify the existing MCS low pressure reactor trip.
Tne changes to the main steam line NRVs involve the addition of a cuick closing (3-5 seconds) stored energy actaator on each of the four NRV:. The.NRVs will r e-place turbine stop valves (TSV) as the ouick acting main steam isolation valves.
L Each main steam line (MSL) will be fittec with three pressure switenes (3 channels per steam line; one pr,are switch per channel) upstream of the NRV in that line. Each pressure switch operates a Train A and a Train B logic relay. Contacts frorn these relays will be combined to fom a 2-out-of-3 trip-logic te automatic:ily generate a NRY trip signal. Channel relays will be ens gi:ed to actuate a trip.
In addition, the NRV tr'p signal will be used.to generate a reactor trip.
i The changes to the main coolant system pressure reactor trips consists of the j
addition cf a high pressure trip and modifications to tne low pressure trip.
The low pressuri:er pressure reactor trip has been deleted from the plant Technical Specifications.
Instead of a reactor trip on either low pressurizer i
pressure (1 channel; l-out-of-1 logic) or ma n coolant system low pressure (1 channel; l-out. of-l logic), a reactor trip will now occur on MCS low pressure only (3 channels; 2-out-of-3 trip logic).
In addition, a MCS high pressure trip has been added (3 channels; 2-out-of-3 trip logic). The same pressure s
s transmitters are used to generate both the high and low pressure trip signals.
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__ 2,0 EVALUATION The NRV trip circuitry consists of three redundant channels of pressure switches per steam line.
Each channel is powered from a separate 125 Vdc battery backed emergency bus.
Each pressuie switch operate:, a Train A and a Train B channel relay. Contacts ' rom these channel relays are combined to for n Train A and Train B *iRV trip signals (2-out-of-3 energize to actuate trip logic:
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logic train (A and B) is powered from a separate 125 Vdc battery t,acked emergancy bus. No single power source failtre will prevent a NRV trip when
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required (i.e., only 1 train would be affected).
Each MSL low pressure channel is run in dedicated cond91t freer, the sensor to the control panels in the main contro' room where separation.. thin the cabinets is maintained by either six inches air space or by physical barriers.
the capability to manually initiate a NRV trip t.t the systam level from the control room is provided by a Train A arid a Train B f2V trip selector switen 1
located on the main control board (MCB). No single failure within either the me.ual or automatic NRV trip circuits will prevent a NRV trip by either manual or automatic means. The NRV trip channels and logic are used to provide
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the NRV trip signals ind control room innunciation.
In addition, both the Train A and Train B NRV tr., signals will cause a reactor trip. The Train B f
NRY trip signal is also used to generate a condensate pump trip permissive signal to allow the operator to manually trip the condensate pumps. This is a non-j safety function.
Isolation from the NRV trip circuits is provided by relay 4
coil-contact separation. There are no control functions associated with the NRY trip circuitry, t
l l
'w Train A and Train 5 system level NRV block switches (spring return to neutrC,) are located on the main control board.
These switches provide the operator the capability to manually bypass the NRV trip function pro-vided a NRV block permissive logic (2-out-of-3 low steam line pressure) is satisfied. This bypass will be automatically removed when system pressure increases above the block pemisshe setpoint. The NRY trip block may be de-feated by placing the NRV block switch in the reset position.
Block pemissive and NRV bypass annunciation are provided as noted below. There are no other overrides or bypasses associated with this, trip function.
Each individual NRV may be " locked open" by placing its control switch in the open position. This condition is alamed in the control room.
In addition, tnc capability is provided to test each NRV during power operation by depressing its Train A or S test pushbutton switch which allows the valve to start closing. Once the NRV reaches tk 10% closed position, limit switch contacts in the NRV test circuit open to prevent further closure.
The following annunciation / indication is provided in the control mom for the NRY trip function:
1.
Two sets of' NRV position indication lights in the main control room (MCB and graphic display),
2.
NRV " Trouble" alarm (loss of power,' low accumulator pressure, low hydraulic fluid pressure, and low hydraulic fluid level),
3.
Train A and B " Channel Trip" alams, 4
Train A and B "NRV Trip" alams, 5.
Train A and B " Control Switch cocked Open" alams, 6.
"NRV House Ambient Temperature Low" alar =, and
s l 7.
Train A and B " Block Permit" and " System Bypar," 71 arms.
Inis infonnation is sufficient to provide the operator with the current states of the NRV trip system.
The MSL low pressure charnels will be tested each shift (channel check; NRV status lights), monthly (channel functional test),. and at refueling inter-vals (channel calibration). The automatic ac Jation logic will be tested quarterly for, each train.
The licensee has stated that the NRV trip 'ise comprised of safety grade environ-mentally qualified components with the exception of the HGA relays which are control grade (HGA relays are used in the NRV control circuits) and the MSL pressure switches wt 'ch are " vendor qualified" and are currently undergoing
" comprehensive safety grade qualification." Ine licensee has ordered safety grade relays and has committed to install them in place of the HGA relays during the first outage following shipment. Operation with the HGA relays installed until this outage is justified since the NRV trip function pro-vides an improvement in overall plant safety and since the licensee has stated' that the HGAs are "high quality" relays.
The main coolant system (MCS) high and low pressure reactor trips each con-sist of three redundant MCS pressure channels (1 channel per main coolant loop T, 2, and 3).
Ea'ch channel receiv'es power via a separate transformer, each of which is fed from the same 120 Vac vital bus. This channel power source configuration is the same as used for other reactor trip functions.
The same pressure transmitters are used to provide both the high and low pressure signals (i.e., each transmitter feeds 2 bistables; one high pressure and one low pressure). There is one pair cf relays (Train A and Train B) for each MOS high and low pressure channel. These relays operate contacts
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. combined to form a 2-out-of-3 de-energize to actuate trip logic. Thus, loss of the 120 Vac emergency bu:. will cause both high and low pressure reactor trips. The failure of a channe; transformer will cause a one-ha'f trip on both MCS high and low pressure. The' Train A and Train B trip logic are powered from different 125 Ve,c vital buses. f6 single failure at the channel or system level can prevent a MOS high or, low pressure reactor trip when required.
Each pressure transmitter is located in a different area within the contain-mer.t. Each ICS high pressure and low pres'sure channel is run in dedicated conduit through a separate penetration to the main control board, thus physical separation is maint.ined.
The MCS high and low pressure trip circuits are,used to pNede reactor trip and co > trol room annunciation and indication functions.
Other indication /
annunciation is provided through isolation devices.
The fcilowing anrmnciation/ indication is provided in the main control room for the nigh and low pressure reacttr trip functions:
1.
Train A and B - 1/3 high pressure trip (channel t 'p) 2.
Train A and S - 1/3 low pressure trip (channel trip) 3.
Train A and B - loss of power 4
Train A and B - hi.gh pressure reacto'r trip 5.
Train A and B - low pressure reactor trip 6.
Channel in test This information is sufficient to provide the operator with the current status of the iCS high and low pressure reactor trips.
There are no blocks, overrides, or bypesses associated with either the high or 1
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- low pres ure trips.
Control roo.T. annunciation is pr.~ided indicating wnen a M:S hign/ low pressure channel is being tested.
Surv0!Ilance requirements fcr both the high and low MOS pressure reactor trips consist of channel checks (once per s ift), channel functional tests (once per month), and channel calibrations (once per refueling outage). Tne CS hig'h and low pressure Train A and Train E reactor trip logic will be tested during each start-up if not perfomed within the previous 7 days.
The licensee has stated that the MOS pressure reactor trips utilize environ-mentally qualified components with the exception of some HGA relays. As with the HGA relays used in the NRV trip circuits, the licensee has committed to replace these relays with a qualified replacement.
Operation with tne HGA relays is justified for the short term since the MOS high and low pressure trips represent an improvement in overall plant safety and since the licensee has stated that the HGAs are "high quality" relays.
In addition, the licensee has stated that the pressurizcr low pressure reactor t.*:p will remain functional altnough it has been deleted from the plant Tect' 4;al Soecifications.
3.0 CONCLUSION
Based upon our 're view of the licensees submittal of Farch 26, 1931 and sub-sequent telecons of June 8 and 18,1981, we conclude that the NRV trip modifi-cation, satisfies the requirements of IEEE Standard 279-1971; and therefore, are acceptable with the exception of the HGA relays which' are control grade.
The licensee has committed to replace these relays with qualified relays as mentioned above and the proposed schedule for replacement is acceptable.
..The proposed Technical Specification changes associated with the NRV trip submitted in the March 26, 1981 letter are correct.
Our review of the MCS high and low pressure reactor trips has concluded that these trip functions satisfy the requirements of.IEEE Standard 279-1971,
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with the exception of the control grade HGA relays, and that the associated Technical Specification changes are correct, and therefore, are acceptabl'.
The licensee has committed to replace the HGA relays with -qualified safety-grade relays and the proposed schedule for replacement. is acceptab!a.
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MISCELLANEOUS CHANGES 1.
A plant modification was implemented during the (c re XV refueling to add dual contactors to SI MOV-46 (HPSI flow control valve) in order to permit throttling of this valve from the control room without mtr.ual action to first connect power to the valve by closing the breaker at the motor control center. We had previously required that this valve be placed in its normal position with power removed from the motor. Now the motor is equipped with dual motor starters in series and two independent control. circuits and switches. Both interrupting devices must be energized by control switches to change valve position from its 7:r rmal position. Position indication for the valve is provided at the safety injection panel in the control room.
We previously approved identical control schemes for other ECCS MOVs in Amendment No. 52, which was issued on November 14, 1978.
We therefore conclude that the controls for this MOV meet the single failure criterion and the requirements for Class lE systems and are therefore acceptable.
f 2.
On pages IV a new subheading under 3/4.3.3 has been added to correct an editorial error from a previous amendment.
SUMMARY
OF FINDINGS l
From our review of the material submitted by the licensee on the Core XV reload, j
including the modifications to the Auxiliary Feedwater System and Non-Return Valves, we find:
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F o A.
The mechanical design of the fuel, the nuclear and thermal-hydraulic analysis, and the analyses of accidents and transients are acceptable.
B.
The modifications to the Non-Return Valves (NRV) and the Auxiliary i
Feedwater System ( AFW) are acceptable.
These modi fications, as described in this Safety Evaluation, are changes to automate the NRVs and to improve the capabilities of the AFW system by the addition of two notor-driven pumps and associated piping.
C.
The proposed Technical Soe ifications, which implement the NRV and AFW changes, and which modify the reactor core operational limits, are acceptable.
Environmental Consideration We have determined that the amendmentdoes 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 and pursuant to 10 CFR 51.5(d)(4) that an environmental impact statement or negative declaration and environmental impact appraisal need nct be prepared in connection with the issuance of this amendment.
69 -
I Conclusion We have concluded, based on the consideratiaa-discussed above, that:'(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable 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 con-ducted 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.
Date: July 22,1981 4
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