ML20148G092

Safety Evaluation Supporting Amend 21 to License DPR-3

ML20148G092
Person / Time
Site:	Yankee Rowe
Issue date:	12/04/1975
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20148G075	List: ML20148G072 ML20148G077 ML20148G085 ML20148G092 ML20148G096 ML20148G102
References
NUDOCS 8011070186
Download: ML20148G092 (17)

Text

l b UNITED STATES 4

- NUCLEAR REGUL ATORY COMMISSION p,':

WASHINGTON. D. c. 20555 .:.

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.... :f ra SAFETY EVALUATION BY Tile OFFICE OF NUCLEAR REACTOR REGULATION Hu 'y i

i SUPPORTING AMENDMENT NO.2 1 TO'TACILITY OPERATING LICENSE NO. DPR-3 [

(CifANGE NQ.I q e TO THE TECilNICAL SPECIFICATIONS) )..

h.

YANKEE ATOMIC ELECTRIC COMPANY a

d M

YANKEE NUCLEAR POWER STATION (YANKEE-ROWE) h DOCKET NO. 50-29  !

Introduction )

t By letters dated July 14, 1975 (as supplemented October 10, October 28, November 7 (Proprietary Information appended), November 21, and November 26, 1975]; .luly 8, 1975 (as supplemented November 24, and p~

November 26, 1975); and September 23, 1975, Yankee At omic Electric Company (the licensee) requested changes to the Technical Specifications appended to Facility Operating License No. DPR-3 for the Y ukoc Nuclear Power Station (Yankec-Rowc) . The purpose of the requests is to change the operating limits and to add requirements relating to the Emergency Core Cooling System (ECCS) in the Technical Specifications. The revi-sions to the Technical Specifications are based on an acceptabic evalua-

• tion model that conforms to the requirements of 10 CFR Part 50, 850.46, to permit operation of Yankee-Rowe with Core XII reloaded with new Exxon Nuclear Company (ENC) fuel assemblics and recycled Gulf United d Nuclear Fuels Company (GUNF) fuel assemblies irradiated in the preceding y Core XI. q R

Discussion . l yt The Yankee-Rowc reactor core consists of 76 fuel assemblics, each having a 16x16 array of fuel rods. The Core XII reloaded core utilizes a two- 1 regional configuration with the 40 ENC fuel assemblics located'at the i]

a periphery of the core and the one-cycle exposure GUNF fuel assemblics occupying the interior of the core. j The licensee provided the needed technical information for our review, including a general description of the reload core, detailed mechanical y design data on the reload fuel, the results of the nuclear and thermal- .3 hydraulic evaluation, accident ond. transient analysis in support of the 3 Core XII reload application. Since this is the first application for a {

Yankec-Rowe reload with ENC fuel assemblics (the second ENC reload appli- }

M cation to PWRs following 11. B. Robinson, Unit 2), ENC has provided docu- "

mentation on che ENC, ECCS cooling performance analysis models and computer codes.

801107.0 M 6

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l We have examined the methods employed by the licensee and ENC and concluded that their application to the design and analyses of the Yankec-Rowe reconstituted Core XII is acceptable. Further, from our review of the available reload information, we conclude that it is acceptable for the licensee to proceed with Core XII operation in the manner proposed. Our review and evaluation of the licensee's Core XII reload submittal is discussed in the following sections.

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Evaluation A. Fuel and Mechanical Design The proposed reload core (Core XII) consists of 40 fresh Exxon Nuclear fuel assemblies and 36 one-cycle exposure GUNF fuel assemblies. A comparison of the mechanical designs for the ENC and GUNF fuel assemblies indicates the following differences:

l'. The ENC fuel assembly has an open lattico design which uses stainless steel guide bars for structural support. The GUNF design has a stainless steel shroud around the assembly. 1hc fuel rod pitch in the ENC design has been increased from 0.468 in h to 0.472 inch to compensate for the removal of the shroud.

2. The upper no::le of the ENC fuel assembly design permits  !

removal and reinsertion of fuel rods. p

3. The E5C spacer grid is a stainicss steel and Inconel bimetallic l assembly representative of the manufacturer's standard product L line. -  !

4. The lower no::le of the ENC design is mechanically attached to the guide bars.

5. The ENC fuel pellet dcsign for Yankee-Rowe is the same as all other ENC pellet designs but shorter than the GUNF Core XI i.

pellet design.  !

The licensee has performed proof tests to assess the structural inte-grity of the new fuel as: mbly design. These tests examined the strength of the locking system between the upper no :le assembly and the guide bars and the strengths of the spacer assembly, lower no::le, and upper tie plate. Tests were also conducted to determine the sup-port stiffness oi'the spacer dimples used in design calculations and the functional loading between the spacer grid contact points and the fuel rod cladding. In addition, ENC performed a short duration combined fretting wear and pressure drop test on a prototypical Yankee-Rowe fuel assembly.

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b Several analyses using generically approved codes confirmed that :p the ENC fuel assembly design limits were met. These analyses in- Q:)

cluded strain limit calculations, cladding collapse calculations, _ f'" i]

and fuel temperature calculations (including the effects of fuci ......=

densification). No found that the ENC engineering methods, design "1" limits, and results wore satisfactory. ~,"

Nino one-cyclo exposuro Core XI GUNF fuel assemblics to be reinserted in Core XII were visually inspected ind no apparent rod bow was observed.

Examination of similarly designed GUNF BWR and PWR fuel, irradiated between approximately 5000 to 15,000 MWD /MTU, also indicated no apparent rod bow. In addition, the ENC and GUNF fuel designs have ,

a larger thickness to diamotor ratio than the standard Westinghouse ISx15 fuci design (0.0066 compared to 0.0057). The distance between gd the spacer grids of the ENC and GUNF fuel is considerably smaller Q than that of Westinghouse fuel (14.7 compared to 26.2 inches) . [ p These differences should result in decreased rod bow compared to P d that measured in the Westinghouse fuel assemblics.

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Yankec-Rowe is one of the first plants to use ENC, PWR typo reload ,

assoublics. ENC's only other operating experience to date on PWR 3 fuel has been with two. lead assemblics in the Rochester Gas and Electric Corporation's R. E. Ginna reactor which were inspected .q 3

after one cycle and had no leakers. Responding to our question

- regarding surveillance of ENC fuel assumblics, the licensoc has committed to an inspection program of at least 5 of the 40 Core XII,. ,

ENC reload assemblics during cach of the next tva refueling outages.

Specific components to be examined will include top and bottom no :les, ,

guide bars, spacer grids, and fuel tubes. If irradiation of the Core XII fuel will be continued beyond the next two cycles, the sur-veillance progran will'also be continued through the highest burnup '

achieved. J From our review of the information provided by the licensco on the Core XII reload submittal, we have determined.

1. The ENC fuel rod mechanical design is coupatibic with the pre-viously approved GUNF Core XI fuel design and provides acceptabic engineering safety margins.

2. The analysis performed acceptably accounts for the effects of fuel densification, cladding collapse, and cladding strain.

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3. The results of the out-of-pilo proof tests reported varify the adequacy of the fuel mechanical design.

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K 4.1The surveillanco program Lfor the ENC Coro XI_1 fuel -hssemblics

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isl acceptable' for evaluation of the fuel performanco. t n

h'o,: therefore, concludo that, from a mechanical ' design standpoint, .....l=

l reloading .of 'the core with the previously approved GUNF fuel assem-blies in Core. Region A and with the' newh ENC- l ded fuel assemblics Coro XII in' Core

. Region .B and operation of Yankco-Rowe with t e ro oa l -

are acceptablo.

• Nuclear Design a

' . B.

The Core XII reload configuration departs from tho throc-zone pattern I[

l lused in preceding Yankee-Rowe cores. . Coro XII.includos.substan . ~

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tially more than the usual amount ofCore fresh fuci, nr.d XII also E

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.. th higher boron concentration for a-lower t averago t This burnup.has a .

.brium reload coro because there is.lcss plutoniun prosen .blics in- e

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. occurs because of the absence of two-cycle exposure assem [

used in the accident analysis are' chosen in a conservativo mnnner s i::

for each analysis. The range of data calculated for Core XII, thereforc. lie within that used for the accident analysis [

9 The licensee's calculation of control group worths. for Core XII', indi- '

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" cates that there is a substantial excess margin overA a7.5% 1%Ap design '

uncer-shutdown margin allowance throughout the cycle life.

tainty allowance has been included in the calculaticn of rod worth.

Comparison with Core XI' calculations ofStartup measured worths indicates that monsurements this uncertainty allowance is conservative.

will provide additional verification that the shutdown margin will-be maintained throughout Cycle XII operation.

The nucicar calculations for Cyclo X11 were performed by the licensoc ,

using'the same calculational methods employed for Core XI.which we have previously found acceptable.

Peak linear heat gunerat'on rates (UlGR) are restricted as a function of Cycic XII exposure as indicated in the section on EC which limit the reactor power as follows:

LOCA limit UlGR i: Allowable Fraction of Power = I:ull power UIGR

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m The full power LilGR is, the product;of

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1. .the : core average LHGif at full power- (4.34' kW/ft),

I I 2. .the mea'sured total . nuclear. peaking factor, Ph,^ .(determine'd every i j

,1000 equivalent full power. hours),. T : f@

-3.. a. factor which allows for.the increase in peaking factor if f the

' rods are'-inserted to their Iinsortion limit,(1.0 for unrodded measurecentof.F(), ))]

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~4. .a factor.which allows. for the maximum increase in peaking factor N which could o'ccur from zenon redistribution (1.10),

5. a flux peaking augmentation- factor which allows for axial gaps in the fuel from densification- (axially variable),- =a

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6. a measurement. uncertainty allowance (1.05),.

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7. an allowance for. power level uncertainty (1.03), ;l q

, 8. an allowance for engineering uncertainty (1.04), and s R

3 9. an allowance for the pellet stack height shortening from fuel 'l

'densification (1.009) . ;j

q Items 3 through 9 impose ~a total uncertainty penalty in excess of 30t. .h This is conservative because it is extremely unlikely that all of 1

-these factors.would have their combined maximum adverse.offect at any a given time. The proposed Technical Specification D.2.C will' effectively 5 j limit the reactor power to'a'icyc1 consistent with the LHGR used in the LOCA analysis and it will do so in a very conservative manner. p, We, therefore, find it acceptable. ""

c-C. Thermal and Hydraulic Design _

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.he features in the design of the new ENC fuel assemblics that might ,

affect the thermal and hydraulic performance of Core XII involve the '

change of the structural suppprts and a slight increase in the clearance  ;.c.

between fuel rods. The GUNF fuel has perforated shrouds that provide structural supports for the fuel rod assemblics. The new ENC fuel is  : 5 of an open lattico design that utilizes eight non-fueled tie rods in

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lieu of shrouds for providing the structural supports for the fuel rod E assemblies. The new ENC ~ fuel assemblics and the one-cycle exposure GUNF fuel

",. assemblics in Core XII have the same fuel rod dimensions. However, the clearance between fuel rods is slightly larger in the new ENC fuel to compensate for the absence of the-shrouds.

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The hydraulic resistance of the new ENC fuel,assemblics is less than that of the one-cycle exposure GUNF assemblics in Core XII because of the ==5. .

absence of shrouds (resulting in decreased' wetted surface) and the s == s reduced. contraction / expansion lossos across the spacers. As a result,  !~

the average flow in the new ENC assemblics is higher than in the GUNF ..;

. assemblies. In our review, we have also considered potential flow ,

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! diversion between the different fuel types at the spacer clevations *

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because of the differences in hydraulic resistance. We find that the  ! ..

D l mismatch in' hydraulic characteristics in Core XII is less pronounced I (therefore, less flow diversion) than existed in Core X which contained  ;

stainicss steel and zircaloy. clad fuel. assemblics of distinctly dif- "

. forent designs.

6 The licensec's previous thermal-hydraulic analysis for Core X was based

.on Cat II calculations. We found the results of those conserva-

tive calculations acceptable ta support operation with Core X (con-l taining -ircaloy clad fuel with perforated shrouds) at a peak linear ,

4 heat generation rate of 12.2 kW/ft with a hot channel enthalpy risc BE factor of 2.24. For Core XI which consisted of. 72 zircaloy clad assemblies and four stainicss steel clad assemblies, the licensee's thermal and hydraulic calculations were based on Cobra-3C. Those calculaticas showed that the DNB ratios predicted for a range of abnornal operating transients in Core XI all cxceeded 2.0 for design j hot channel conditions of 12.9 kW/ft and F6H of 1.81.

The licensce's LOCA analysis for Core XII shows that the limiting hot N channel conditions will be lower than the design conditions used for j the thermal and hydraulic design of Core X or Core XI. Because the ,

operating limits for Core XII are more restrictivo (due to the restric- ,

tions imposed by the ECCS coro cooling performance analysis) than  !:

those previously approved by us for Core X and Core XI,' we find that a the thermal and hydraulic design of Core XII to be acceptable.

1 D. Accident and Transient Analysis

}

] 1. ECCS Cooling Performance (LOCA) Analysis I

a. Evaluation Model i

The licensee.has evaluated through ENC the Yankec-Rowc ECCS cooling performance using a calculational model that conforms to the requirements of 10 CFR Part 50, 550.46.

The calculational model used by ENC for Yankee-Rowe is similar to the approved H. B. Robinson ECCS performanco evaluation l-codel addressed in the staff's Safety Evaluation of September 11, ,

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s. ;;. .j 7-1975, and its supplement. Time step and nodalization studies have been submitted in support of the model. We have reviewed the use of the H. B. Robinson model for the Yankoc-Rowe ECCS performance evaluation with respect to the ' differences in plant design, particularly the shorter s core, the thinner fuel rods, and the different accumulator I arrangement in the Yankee-Rowc facility. We have determined  :

that the H. B. Robinson model conservatively accommodates those differences and therefore conclude that its applica-tion to the Yankee-Rowe plant is acceptabic.

• i

b. Break Spectrum Using the acceptabic evaluation model described in the preceding -

section, the licensec provided in the November 26, 1975 sub-mittal the results of the analysis of a limited break spectrum [the largest double-ended cold leg (DECL)] guillotine []

break and an equivalent DECL split break. This analysis identified the DECL split break at the pump discharge is to be the lJ

' U most limiting break, with a calculated peak clad temperature d 4

of 10830F, well below the acceptable limit of 22000F as spe- d cifi de in 10 CFR Part 50, S50.46(b). In addition, the maxi- l mam local metal / water reaction of less than 2.2% and the d total core wide metal / water reaction of less than 1% were l within the allowable limits of 17% and 1%, respective 1), d Based on this analysis the licensee proposed to limit the ,. , .]

peak linear heat generation rote (LHGR) to 9 kW/ft for j

'q impicmenting the analysis results.

D he have compared the results of the breaks analyzed for  ;

Yankee-Rowe wit'h the results of analyses that we have pre- 4 viously reviewed and approved for various other facilities, '

particularly that which was performed with the approved ESC  ; y model for the H. B. Robinson facility. Based on this compari-son, we have cencluded that the licensee has provided acceptabic bounding calculations, has identified the most limiting break, and has proposed acceptable Technical Specification limits q for the LHGR which assure operation of Yankee-Rowe with ^

Core XII in compliance with 10 CFR Part 50, 550.46.

To improve the effectiveness of the Yankee-Rowe Technical Specifications, we have also included in Section D.2.e(4) y of Appendix A a requirement to maintain the accumulator water level at a minimum of 700 fta, as proposed in the licensee's September 23, 1975 submittal. The licensee has used this value as an input in the ECCS cooling performance evaluation for Core XII which we find acceptable.

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c. ECCS Containment Pressure Evaluation Appendix K to 10 CFR Part 50 requires that the effect of operation of all the installed pressure reducing systems I?.::

For. E and processes be included in the ECCS evaluation.

the evaluation it is conservative to minimize the contain- ~g ment pressure since this will increase the resistance to  :

steam flow in the reactor coolant loops and reduce the -

reflood rate in the cor'c. Following a loss-of-coolant accident (LOCA), the pressure in the containment will be increased by the addi lon of steam and water from the primary reactor system into the containment atecsphere.

After initial blowdown, heat flow to the ECCS water from the core, primary metal structures, and steam generators will produce additional steam. This steam together with z.:

any ECCS water spilled from the primary system will flow {.q :

through the postulated break into the containment. This 3 energy will be released to the containment during both the blowdown and l'ater ECCS operation phases; i.e., reflood and

., post-reflood phases.

. Energy removal occurs within the containment by several means. Steam condensation on the containment shc11 and internal structures serves as a passive energy heat sink that becomes effective early in the blowdown transient. .j Yankec-Rowe has a bare steel > containment shell capable ]

of transferring heat generated during a LOCA from inside the containment to the atmosphere outside the containment. 1

No other systems are needed for heat removal, h* hon the y energy removal rate exceeds the rate of energy addition from j the primary system, the containment pressure will decrease ,

W from its maximum value, uj The ECCS containment pressure calculations for Yankee-Rowe -

were done using the ENC ECCS cvaluation model, h'c have reviewed the ENC ECCS cvaluation model and discussed it in our Safety Evaluation dated September 11, 1975. We concluded that ENC's containment pressure model was acceptabic for the ECCS cooling performance evaluation. We required, however, that justification of' the plant-dependent input paramotors "'

used in the analysis be submitted for our review of each plant.

The licensee submitted justification for the containment input data for Yankee-Rowe by lotter dated November 26, 1974.

The licensee has reevaluated the containment not free volume W

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and the passive heat sinks with regard to'the conservatism for the ECCS performance analysis. This reevaluation was based on measurements within the as-built containment to which additional margin was added. There are no active heat removal systems installed in the Yankco-Rowe contain-ment.

ENC has used a constant value of 11.8 psig in the core heatup calculations. We have compared this value with our indepen-dent calculation of the containment pressure as a function of time based on ENC's calculated mass and energy release data and the containment input data for the Yankee-Rowe plant.

This analysis demonstrates the value of 11.8 psia used by ENC to be conservative to a time of 115 seconds after the accident. (Peak cladding temperature is predicted to occur before 115 seconds.)

h'c have concluded that the containmer 'tre analysis for Yankee-Rowe ir reasonably conserv- refore, con-forms with Appendix K to 10 CFR Paz. ...

d. Single Failure Criterion Appendix K to 10 CFR Part 50 of the Commission's regulations requires that the combination of ECCS subsystems to be assumed operational shall be those availabic after the mast damaging single failure of ECCS equipment has occurred. The worst -

single failure of ECCS equipment which would minimize the ECCSavailable to cool the core was identified by the licensee -

as the failure of one of the three emergency diesels to start.

A review of the Yankee-Rowe piping and instrumentation diagrams indicated that the inadvertent actuation of specific motor-operated valves could affect the appropriate single failure assumptions. We had identified the following motor- .

operated valves which did not satisfy the single failure criterion.

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MOV# Component Function Failure Mode SI-MOV-1 Accumulator isolation Closure would cause loss of accumulator injection to RCS SI-MOV-22 Isolation valves from SI he'ader to Closure would enuse SI-MOV-23 cold Icg injection lines reduction of ECC SI-MOV-24 flow to RCS SI-MOV-25 SI-MOV-4 Crossover from LPSI pump discharge Closure of the valve to HPSI pump suction would prevent boosting of SI flow to HPSI pumps for small break SI-MOV-46 Flow control from HPSI pumps Closure of the valve

. would climinate HPSI flow to RCS SI-MOV-49 HPSI test /recire line valve (mini- Fail closed-overheat flow) pumps CH-MOV-522 Isolation valve between charging Valve should remain pump discharge and LDSI discharge closed header h..

CH-MOV-523 Isolation ialve from charging pump. Closure of either CH-MOV-524 discharge headcr to hot leg of valve during long term l loop 4 (provide normal charging recirculation would }

function and hot leg injection for prevent hot leg injec-long-term recirculation) . tion CS-MOV-536 Isolation valves from SI header to Closure would reduce CS-MOV-537 cold Icg injection lines ECC flow to RCS CS-MOV-538  ;

CS-MOV-539 C3-MOV-532 LPSI test / recirculation line valve Opening of valve would (full flow) reduce LPSI flow to RCS CS-MOV-533 LPSI pump discharge isolation valve Closure would prevent CS-MOV-535 LPSI flow to RCS 4

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MOV# Component Function Failure Mode CS-M3V-301 RCS loop isolation valves Closure would isolate CS-MOV-302 a RCS loop CS-MOV-309 5

CS-MOV-310 p CS-MOV-318 CS-MOV-319 CS-MOV-325 CS-MOV-326 SI-SV-56 Accumulator nitrogen pilot relief Premature nitrogen SI-SV-5 7 valves relief would reduce accumulator injection The licensee has reviewed the existing Emergency Operating Procedures and the consequences of the identified single failures and has proposed the following.

(1) The present Emergency Operating Procedures will be modified to eliminate the ECCS " cutback" mode of operation.

(2) During power operation, a.c. power will be removed from the following motor-operated valves with the valves in their norna11y opened position by removal of the circuit breaker frem the motor control center:

SI-MOV-1, SI-MOV-4, SI-MOV-22, SI-MOV-23, SI-MOV-24, .

SI-MOV-15, SI-MOV-46, and SI-MOV-49.

(3) Valves CH-MOV-522, Cll-MOV-523, and Cil-MOV-524 will be rewired to a common circuit breaker and a second series breaker will be installed. During power operation, a.c. power will be removed from Cll-MOV-522, CH-MOV-523, and Cil-MOV-524 with Cll-MOV-522 in its normally closed position and Cil-MOV-523 and Cil-MOV-524 in their normally opened position by opening both of the series breakers. D (4) Using EICSB 13 ranch Technical Position 18 as guidance, position indication for valves Cil-MOV-522, Cll-MOV-523, and CH-MOV-524 will be provided by rewiring, making the position indication features independent of breaker position. In addition, valve position indication for Cll-MOV-522 will be provided by operator monitoring of flow through flow indicat6r..FI-2 and valves CH-MOV-523 and CH-MOV-524 will be pinced under periodic operator surveillance to verify their proper alignment.

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(5) During power operation, a.c. power will be removed from the following motor-operated valves with the valves in thcir normally opened position by discon-necting the power cables as they 1 cave the motor starters: CS-MOV-533, CS-MOV-535, CS-MOV-536, CS-MOV-537, CS-MOV-538, CS-MOV-539, MC-l10V-301, MC-liOV-302, MC-MOV-309, MC-MOV-310, MC-MOV-318, MC-MOV-319, MC -MOV-32 5, MC-MOV-326.

(6) During power operation, a.c. power will be removed from motor-operated valve CS-MOV-532 with the valve in its normally closed position by disconnecting the power cables as they Icave the motor starter. In addition, this valve will be permitted to be opened to provide the capability for mixing of the coolant in the Safety Injection tank on a quarterly basis for a period of approximately thirty minutes.

(7) The control circuitry to each of the accumulator pilot operated relief valves, SI-SV-56 and SI-SV-57, will be modified to include level switch contacts in each of the conductors to and from the valve solenoid operaters. This will preclude a single failure from opening the accumulator relief valves before termination of the injection mode of operation.

(8) As noted in item 3 above, CH-MOV-523 and CH-MOV-524 will be deactivated in their normally opened position.

This will assure that normal charging flow and long >

term hot leg rceirculation will be provided. Since hot leg injection during the inj action mode of opera- l tion hr.s not been justified as an acceptable procedure, the licensee has proposed to prevent hot leg injection during injection mode operation by tripping the charging pumps upon receipt of a safety injection j actuation signal. Safety injection actuation is com-prised of systems "A" and "B" which together provide redundant actuation signals to all safety injection equipment. In order to provide the required tripping function and meet the single failure criterion, the-licensee has proposed to provide two series contactors  !

in the power supply circuitry to each charging pump. j".

The "A" safety injection signal initiates a trip to one contactor while the "B" safety injection signal l initiates a trip to the other contactor. This modi-fication will assure that charging pump flow will be provided to the hot 1 cgs only when r'equired for normal charging and long term recirculation, and prevent hot ,

Icg injection during the injection phase of operation. I l'

(9) The onsite emergency power system consists of three dicsc1 generator busses, each powering a high pres-sure and a low pressure safety injection pump. The e+

ECCS analysis demonstrates that the loss of one diesel

- and its associated equipment can be tolerated without exceeding Appendix K requirements. In our review, however, we have determined that two of the three dicsc1 generator busses are not independent. Busses 1 and 3 are each a normal and an alternate source of power for a swing bus. Yankee-Rowe has two swing bussos which power redundant ECCS valve trains and are cach connected to busses 1 and 3 through an automatic transfer switch. The automatic transfer capability is not required to meet ECCS acceptance criteria. The flexibility afforded by these swing busses is far out-weighed by the reality that a singic failure can com-promise two of the three safety trains. b'e have required that this aspect of the onsite emergency power system design meet the single failure criterion

> and conform to the recommendations of Regulatory Guide 1.6. To this end, the licensee has prcposed to rack out and lock the alternate supply breakers and within 60 days fron the date of startup with Core XII proposo and provide mechanical interlocks so that the normal and alternate supply breakers cannot be closed simultaneously. Also, we have required that the auto-matic transfer switch feature be removed and that the -

switch be operated manually. We find that these modi-fications will bring the onsite emergency power system  ;

into conformance with the single failure critorion and  ;

are therefore acceptabic. ,

Conclusions The staff has reviewed the licensce's proposals for satis-fying the single failure criterion and has found them to be acceptabic, h'ith the modifications, as discussed above, the plant will satisfy the requirements of Appendix K to 10 CFR Part 50 of the Commission's regulations. .

L

e. Long Term Boron Concentration Builduo h h'e have reviewed the licensce's propcsed emergency operating procedures and the systems designed for preventing excessive boric acid buildup in the reactor vessel during the post-LOCA long term cooling period. The licensee has proposed for the first 20 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a LOCA to inject borated water i

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a from the c'ontainment sump;into the RCS cold icgsjby means ;g <

of: the purification pumps. After 20 to-24 hours,(the ECCS (. .....s:c;'

will be' realigned and the borated solution will- be injected.

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simultaneously into the, hot.and. cold Icgs'. uThis will be EL: -

accomplished by diverting a portion of the' purification --

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' pump output into the' suction of the charging pumps which .

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• will: deliver the flow to the hot. legs. A hot leg injection- i53 flow of.25.2 gpm is required to. control the boric acid F concentration in the core. ,

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n'e have reviewed the procedure and ' concluded ' . that it. M

.wil1< satisfactorily maintain the concentration of boric  ;

acid in the core below the solubility linit provided that:

.. .(1) . . Power is. disconnected. to motor-operated valves; 5E

~'CH-MOV-523 and Cll-MOV-524 with'the valves in their ,

normally opened position to.' assure:the' delivery of R the required hot Icg injectant (See Section D.1.d =

{]

above),'and, j...,

(2) The licensee utili:csexisting flow r.ctering instru- d mentation for measuring and. controlling the hot leg  :.j recirculation flow to' assure that the hot Icg recirco-  ;

lation flow required to control the boric acid con- "i contration in the core is provided. }

~;

f. Submerged Valves ' '

The licensco has submitted an analysis on July 8,1975, which  !

reviewed the 'lankee-Rowc equipment arrangement. h'c have con- [ f cluded that no ' valve motors within containment required for  ;

q ECCS operation or long-term core cooling ui' ; become subnerged y following a 1.0CA. y Conclusions' 9 Based on our review, we conclude that:

2; r .

1). The ECCS cooling performance (LOCA) analysis submitted by the j-'" ";

licenseo is in conformance with the requirements of Appendix K to.10 CFR Part 50. Additional analyses will be submitted by .]

the licensee which will confirm that the trends predicted for j the H. B. Robinson plant are appropriate for reference by D Yankec-Rowe, and that the doubic-ended cold leg split break .

is the most limiting break size for Yankcc-Rowe, du, 5, ?

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4

2) The ECCS cooling performance conforms to the peak clad temperature and maximum oxidation and hydrogen generation criteria of 10 CFR Part 50, 650.46(b) .

3) ECCS cooling performance will be adequate despite any postulated failure of a singic active component.

4) Adequate systems and procedures exist to provide long term cooling to the reactor vessel.

2. Control Rod Ejection Accident For the reference cycle (Cycle XI) a rod ejection analysis was performed using the most limiting parameters during the core life. The input parameters are more favorable for the reload cycle (Core XII) except for the delayed neutron fraction, which is slightly lower. Ilowever, using the delayed neutron fraction to compute the ejected rod worth for the limiting case, the value for the reference cycle is 0.86 dollars while for the reload cycle the value is 0.46 dollars. Vierefore, the results for the reload cycle will be bounded by those for the reference cycle and are acceptabic.

3. Control Rod Drop Incident The licensec's bounding analysis of the c;ntrol rod drop incident indicates that dr. rage would not result from this incident even if no drop in core power were assumed. We find that the licensce's analysis and results are acceptabic. '

4. Control Rod Withdrawal Incident, Boron Dilution Incident, Isolated Loop Startup Incident, Loss of Load Incident, Loss of Feedwater Flow Incident, Loss of Coolant Flow Incident, Steam Line Rupture Accident, and Steam Generator Tube Rupture /secident Transient and accident analyses were performed for the reference cycle (Core XI) using the most limiting parameters during the '

core life. For the reload cycle, the input parameters are more favorabic than for the reference cycle. Therefore, the results for the reload cycle will be bounded by those for the reference cycle. We find this acceptabic.

5. Other Accidents and Transients The remaining accidents and transients in the licensco's FSAR are not affected by the proposed core design changes and therefore the previous occeptabic results still apply, p i

i

Summary of Findings .

From our review of the material submitted by the licensee on the Core XII reload, including the ECCS cooling performance evaluation, we find:

1. The mechanical design of the new ENC fuel, the nuclear and thermal-hydraulic analyses, and the analyses of accidents and transients are acceptabl..

2. The ECCS cooling performance for Core XII has been calculated with an approved evaluation model in conformity with Appendix K and meets the acceptance criteria in 10 CFR Part 50, 550.46(b).

3. The modifications to the ECCS to preclude single failures and to prevent boron precipitation during the long term core cooling phase following a LOCA are acceptable. The modifi-cations as described in this Safety Evaluation must be com-picted before proceeding with power ascension following com-pletion of the power physics testing.

4. The proposed Technical Specifications, implementing the ECCS cooling performance evaluation results, provide acceptable limits (these limits are more severe than the restrictions in the Commission's December 27, 1974 Order for Modification of License, which they supersede) for the safe operation of Yankee-Rowe with Core XII.

Conclusions h'c have concluded, based on the considerations discussed above, that:

(1) there is reasonable assorance that the health and safety of the ti public will not be endangered by operation in the proposed manner, and a (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.

, l Date: 3 ,, ' ,;75 4

4 g ..

i -

C. .

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REFERENCES

" Order for Modification of License", letter sent to Yankee Atomic Electric Company (YAEC) from Robert A. Purple dated December 27, 1974.

Letter from D. E. Vandenburgh to Robert A. Purple dated July 8, 1975, submitting YAEC's Proposed Change No. 117, Supplement No. 3.

Letter from D. E. Vandenburgh to Robert A. Purple dated November 24, 1975, submittiny, YAEC's Proposed Change No. 117, Supplement No. 5.

Letter from D. E. Vandenburgh to Robert A. Purple dated November 26, 1975, submitting Proposed Change No. 117, Supplement No. 6.

Letter from W. P. Johnson to Robert A. Purple dated July 14, 1975, submitting Proposed Change No. 125. _

Letter from D. E. Vandenburgh to USNRC dated October 10, 1975, submitting Proposed Change No. 125, Supplement No. 1.

Letter from D. E. Vandenburgh to the Office of Nuclear Reactor Regulation dated October 28, 1975, submitting Proposed Change No. 125, Supplement No. 2.

Letter from D. E. Vandenburgh to the Office of Nuc1 car Reactor Regulation dated November 7, 1975, submitting Proposed Change No. 125, Supplement No. 3 (Proprietary Information appended).

Letter from D. E. Vandenburgh to the Office of Nuclear Reactor Regulation dated Novenber 21, 1975, submitting Proposed Change No. 125, Supplement No 4, Revision 1.

Letter from D. E. Vandenburgh to the Office of Nuclear Reactor Regulation dated November 26, 1975, submitting Proposed Change No. 125, Supplement No. 5.

Letter from D. E. Vandenburgh to Robert A. Purple dated September 23, 1975, submitting Proposed Change No. 132.

Safety Evaluation Report Regarding Review of the Exxon Nuclear Company PWR ECCS Codes and the 11. B. Robinson Reactor ECCS Evaluation Model for Conformance to /<11 Requirements of Appendix K to 10 CFR Part 50 by the Office of Nuclear Reactor Reculation, USNRC, September 12, 1975, and Supplement No. 1 dated Novembe. 1975.

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