ML20085F278

From kanterella
Jump to navigation Jump to search
Change Request 26 to License DPR-4,lowering Settings of PSV-1 & PSV-2 to Max Value of 3,980 Psig
ML20085F278
Person / Time
Site: Saxton File:GPU Nuclear icon.png
Issue date: 04/24/1967
From: Neidig R
SAXTON NUCLEAR EXPERIMENTAL CORP.
To:
Shared Package
ML20083L048 List: ... further results
References
FOIA-91-17 NUDOCS 9110220106
Download: ML20085F278 (6)
v • d • e


Text

- - . .- - - - _ - - - _ -. . . _ _ _ _ _ _ _ _ _

.e 4

0e ,

, g.

SATNN NUCIJAR EU' ERD! ENTAL C0ltPJ:lATION DOCKET NO. 50-146 LICEGE DPR-4 CRANGE RWUIST NO. 26 1.

Applicant hereby sulrtits Chance Request No. 26 in conpliance with paragraph 3.B of license DPR-4 for change of the Technical Specifications to be authorized by the Comission as provided in 10 CFR 50 59.

SAXTGi NUCLEAR EXPERIMENTAL CORPORATION Ey /s/ R. E. Neidig President I

,_, iycyp _20100' l

l FOlr PDR hf3 Rpm:om h

.nm4

4 .

\/ .

j l

April 18, 1967 Docket No. 50-146 DPR-4 Technical Specifications f Change Request No. 26 I

' Page 2 of 2 Pages 3 Safety Consider 3tions 1

.he portion of the loop under consideration is approved by the l Penns,"lvania Departner.t of Industrial Safety under Section I of the ASME Code.

A Pannsylvania Special Number, No. 2162, is assi6ned to the loop since, to

omply with nuclear requirenents, certain components were fabricated under l

the rules of Section VIII of the ASHE Code and the applicable nuclear code case, 1273U.

Tne ecde approved sr.fety valves for the loop are PSV-1 and PSV-2.

These valves are located downstream of the pressure tube, and are now set to actuate at nominal pressures of 4000 psig and (1.23 psig, respectively. These net pressures are based on a preenure tube desQn pressure of 4000 psig. The locatica and setting of these two valves fulfill the requirements of the Code with respect to overpressure protection for the pressure tube and the loop.

Relief valve, PSV-3, which is not required by the Code is installed to provide additional overpressure protection for the loop.

!.owering the settings of PSV-1 and PSV 2 to a naximum value of 3930 poig takes into consideration the pressure drop which occurs across the pressure tube during nornal loop operation. This p.' essure drop is calculated to be 4 na.cinum of 140 psi. The new settings of 39E0 psig will limit the pressure at the tube inlet to a naximum of 103% of the tube design pressure as pemitted by Section VIII of the AS!!E Code.

Iowering the setting of PSV-3 to 4800 psig meets all the requiraaents of Section I of the ASME Code. In addition, this change still allows adequate o;>erating flexibility between the safety valve setting and loop operating pressure, and is consistent with lowering the settings of PSV-1 and PSV-2.

l 4. Health and Safetz It is our conclusion that the health and safety of the public will i not be endangered by this change.

J

I TG:; t;R C.CS

' Gy' ' e' '

U. CALC H0S OTUU L U L'3 H (/jly N me_

TLXA036 TLXC40 (23)AAll8 FC256 fr._ _ r~_

P h,c AGCG i:L FD TDACX SAXTON P:NN APR 11 De C::CW:0LT, ASST DIRECTOR FOR REACTOP OPERATICU3-JIV3 (? HUACYCH '

LIC:. TII:0 UNITED STFor ATEF Div of ATOMIC Comphance EVERGY COMMIS3100 t ,q ,3 0 SFECIFICAT10:0, r& Eb,, E; 200XET NO 50-14 Ga DPD-4, TECH:lIC AL i f f 30Vt.L Ci.u.C_ !:0 QYt.' SUPERCHITIL7I. TECHNOLOGY Fi'OLRAM. (.L::Ci!ANGE f.R.?STED TO CliANGE TECHNICAL SPECIFICATIONS AS FLL1Li,S:

I AL '2 0 :: A'(IllUM I;0 0',' SECTION C.3, ZPRESSURE RELIEF" CHANGE N06:

C.3.A: CliA::0E PGV-1 SETTING FROM 4000 TO 5900, C.3.A; CHANCE PSV-2 SETTIIT FROM 4100 TO 3900 C.3.D; CHANGE PSV-3 SETTI'.A FRO:150C010 4000 FUHF03E:

T: A CET PRESSURES OF SAFETY RELIEF VALVES TO CC:;itCRi' VITH i.F."LICAELE ASME CODES.

SAFETY CCNSIDERATIONS: Ti!E CHANGE TO PLV-1 AllD F3V-2 SETTIGG9 6

it.:(G INTO ACCOUNT MAXIMUM CALCULATED FRICTIO!! PRESCURE DiiOPS 50 T)iAT ASME CODE ALLO'JABLE PRESSURES ON THE LOUEST i:ATED ConPONENT UILL NOT BE EXCEEDED. PSV43 SETTING IS LCtPHE0 TO 4000 SO THAT AT:iE CODE REQUIPEDENTS CAN DE MET EVEN F0:1 ADGOR::AL SITUATIONS AND IC CONSISTENT VITH LOVERING THE SETTIllCS 0F PSV-1 AllD PSV-2.

IT IS OUR CONCLUS10tl THAT T}iE TE ALTH AND SAFETY OF TliE FUDLIC UILL NOT BE IN DANGER BY THIS CHANGE.

APPROVAL IS REQUESTED BY APRIL 17 40YU TO AVOID DELAY OF EXPERICENTAL TESTS PROGRAM i SAXTON NUCLEAR EXPERIMENTAL CORP, R E NEIDIG PRES50-14G DPR-4 N0 16 AEC N0 16 C.3 C.3. A PEV-1 RVPP E0IP C.5. A PSV-2 RQWP E01P C.3.0 PSV-3 TPPP RIPP P3V-1 PSV-2 PSV-3 RIPP M d kl)'~.

xy - < i.w..,+ ,

h' , f,\y, . ~l

'i ' ' 3^ a

\.#

Mi PCV-1 PSV-2 QU 00YUM ll39A 3 4f X

a

-: y

.g0.nh %W ). )

01 l  ??,!

UCALC H3S GTVU .,

=

,t .

//{y g

,,.,G

.L. s

O

-2~

~ . . Variable freqwry Motor Genemtor Set Mcdifications

1. What prevents de reactor coolant pump frun operating frm tM

%D volt bus at high power? Could the underfrequercy mactor trip be made ineffective by opemting the MG set at 63 cycles with the generator breaker (52/VN) open?

2 What are the consequerces of a loss-of-flow accident frun maximum rv. actor power caused by imdvertent op::ning of the IG set gemrator circuit breaker 52/VM? ,

!. Daergercy Core Cooling Controls

1. Could a single failure in the flow totalher keep the pumps frun starting or stop them premturely?
2. Could the injection block switch disable all safety injection?

Can a cingle operator error or a single switch failure disable safety injection?

III. Bemal-Hydraulk Desig Your presentation of the themsl ard hydmulic capability of the core design consists principally of evaluations of steady state and transient Ef3 ratios and fuel temperatures for the hottest core location. A cm-plete assessment of the ccuservatism or safety of the design requires sczne urderstanding of the condition of the entire ora during nomal and transient situations so that we can evaluate the margins available before large ntenbers of fuel rods exceed design limitations. hus, our evaluation of the design nust be based en the ovemil com condition, as well as that of the so called hot spot. A presentation using these considerations arculd be made as follows:

A. Prepare a distribution curve showing the fraction of the core (and number of rods) operuting at the various power levels for design r and overpower conditions, y B. Using the appropriate DNB correlatir.n and the above distribution, detemine the correspording DNB ratios and the statistical nu:nber of fuel rods that could experience DNB.

b o

e

o .. ,

d C. Perfom ar. uncertainty analysis by arbitrarily assuming certain errvrs in major parameters used in calculating the ru.ber of rods experiencing DNB. For example, calculate the numbee of rods with DNB, as a function of possible percentage errors in the WB correlation, power distributions, flow mtes, ard power levels.

IV. Fuel and Materials A. In view of the anticipated high burnups and power levels, is it reasonable to assume that fuel svelling will not cccur at a higher rate than 0.0016 AV/V per 10 20 fs/cc as shown in figure 7 of Beferarce l?

B. What will be the temperature distribution in a fuel element in a high flux position? What will be the temperuture gradient across the cladding, across the gap and in the fuel itself if the density of the fuel is varied between same specified limits?

C. What are the data that support the statement that tube-rvduced Zircaloy would not fail for 7.4 years at the mvi== of 16000 poi expected design stress.

D. What are the applicable ccraposition, dimensicins and inspection specifications for zirronium alloy tubing and the end closure that nake sure that the expected design stresses cannot be exceeded?

E. Under normal conditions stresses in the cladding will be cwessive and strains will be limited by the fuel pellet; but what will be the stress and strain pattem in a sudden depressurization followed by a scrum when fuel temperatures are high and fission gas pressure is

.cffective?

F. Is there experimental data on hydrogen abrorption by the cladding that gp beyond the burnup and power level of Reference 3? What would be the effect of hydrogen absorptico by the cladding c 1 its fatigue pWies, and what would be the combined effect of hydrogen absorption and irrsdiation embrittlement?

V. Test Assemblies Provide (1) ccanparison of the operating conditicns for the test assemblies versus previously approved conditions at 23.5 Mw and (2) safety evaluations for any new steady state or transient conditions.

l 1

1 I -

m. .

4 VI. Pod E3ection Pmtection Prwide a description and drawing of the mechanical st , on the contml mds which are designed to pmvent ejection accident. Irelude an analysis to demonstmte that these stops w:xtid ret fail in the event of a emplete shearing of a contml md nozzle.

VII. @seCalculations By use of the methods end paraneterm of TID-14844, we find that an exclusion radius of at least 300 meters is needed for 35 MW opemtion. Your calcu-lations indicate that the exclusion radius should be et least 270 meters.

Ccmpare the aesunptions used in your calculations to thoce used in TID-14844 Which factors contribute m st to your lower dose estimates?

VIII. Power Escalation Pmgrcm Provide additional dstall on your program for power escalation. Irelude in your discussion, the size of each step in p:wcr iremase (and specific power ircrease), the length of time at each power and the measuxcruds and evaluations required after each step, IX. Dnergercy Cora cooling ,

In our lette" to you dated December 30, 1966, we requested that you con-sider the energency core cooling pmvirions to detemine the need for additional provisions to limit the relese of fission prooucts frun the com. In a3dition, we requested that yw pwfom suicable analyses to detemine the degne to which emergercy core ecoling cast be relied upon to maintair contaiment integrity. Please infom us of the pmgress made thus far ard your schedule for subnittal of the reply.

(

_ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _

Retrieved from "https://kanterella.com/w/index.php?title=ML20085F278&oldid=2739801"