ML20127G701

Forwards SE Accepting Increase in Safety Valve Setpoints,Per 730913 Request to Change TS

ML20127G701
Person / Time
Site:	Monticello
Issue date:	10/02/1973
From:	Skovholt D US ATOMIC ENERGY COMMISSION (AEC)
To:	Mayer L NORTHERN STATES POWER CO.
Shared Package
ML20127G703	List: ML20127G701 ML20127G712
References
NUDOCS 9211170409
Download: ML20127G701 (14)

Text

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- l OCT 2 1973 f.

Docket No. 50-263 s

Northern States Power Company ATTNs Mr. L. O. Meyer, Director of Nuclear Sgport Services-414 Nicollet Ma n Minneapolis, Minnesota 55401 Change No. 10 Gentlement License No, DPR-22 We have reviewed your request (NSP letter dated September 13. 1973)-

to change the Monticeno technical specification set points for the four spring-loaded safety valves on the steam lines at the reactor vessel from two at 1210 psig and two at 1220 peig to four set at 1240 peig and to require four safety valves where three of four insta n ed were required in the past. According to your letter, the increase in safety valve set points to 1240 poig win increase the calculated fuel cycle 2 exposure - threshold, considaring the revised scram reactivity curve and the modi _.ed relief valva response times, by allowing higher system pressures following turbina crip without exceedird the 25 psi GE design margin between peak preauro sad safety valve set point.

A. reanalysis to determine the i., exposure threshold for Monticello fuel cycle 2 is in progress. M.ever, until we have received and evaluated your analysis for. the rematoder of cycle 2, the fuel exposure threshold will be maintained n 1200 MWD /STU for cycle 2 and reactor power will continue to be limited by the fixed control rod inventory established when 1200 MWD /STU exposure level was attained. This is the fuel-exposure threshold calculated prior to resetting safety valves upward to 1240 psig and modifying relief valves to improve response times.

1 i

We have concluded that the increase in safety valve set points from i

1210-1220 to 1240 pois is within acceptable limits to prevent damaging pressure transients and, therefore, do not present significant hasards l

considerations. We have also concluded that there is reasonable assurance j

that the health and safety of the public will not be endangered by-operation of the reactor in the manner proposed.

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9211170409 731002 PDR ADOCK 05000263 P

PDR

Northern Statas Power Company,,,

Accordingly, pursuant to Section 50.59 of 10 CTR Part 50, the Technical Specifications of Provisional Operating License No. DFR-22 are hereby changed by replacing existing pages 16, 20, 21, 23, 24, 25, 26, 85, 118, 119. and 134 with the enclosed revised pages bearing the same numbers.

Our Safety Evaluation is included for your information.

Sincerely, Original Signed bF D. L Ekovbolt Donald J. Skovholt Assistant Director for Operating Reactors Directorate of Licensing

^

Enclosures:

1.

Revised pages as stated above 2.

Safety Evaluation cc w/ enclosures Donald E. Helson, Esquire Harriott Lansing, Esquire VP and GC Assistant City Attorney Northern States Power Company City of St. Paul 414 Micollet Hall 638 City Hall Minneapolis, Minnesota $5401 St. Paul, Minnesota $5102 Gerald Charnoff Warren R. Lawson, X'.D.

1 Shaw, Pittman. Potts. Trowbridge & Madden Secretary & Executive Officer 910 - 17th Street, H. W.

State Department of Health Washington, D. C.

20006 717 Delaware Street, S.E.

Minneapolis, Minn==ota 55440 Howard J. Vogel, Esquire Knittle & Vogel Environmental Library of Minnesota 814 Plour Exchange Building Minneapolis. Minnesota 55415 cc w/ enclosure,cy of NSF ltr dtd 9/13/73:

Steve Gadler, P. E.

Mr. Hans L. Hamester 2120 Carter Avenue ATTN: Joan Sause St. Paul, Minnesota 55108 Office of Radiation Programs Environmental Protection Agency Kan Daugan Room 647A East Tower, Waterside Mall M h - -ata Pollution Control Agency 401 M Street, S. W.

717 Delaware Street. S. E.

Washington, D. C.

20460 Minneapolis, Minnesota 55440 i

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Mr. Gary Williams Federal Activities Branch Environmental Protection Agency 1 N. Wacker Drive, Room 822 Chicago, Illinois 60606 Distribution Adfocket File AEC PDR Local PDR RP Reading Branch Reading JRBuchanan, ORNL TBAbernathy, DTIE VMoore, LIBWR DJSkovholt, L:0R TJCarter, L:0R ACRS (16)

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&ases Continued:

2.1 During transient operation, the heat flux would lag behind the neutron flux due to the inherent heat transfer time constant of the fuel which is B-9 seconds. Also, the limiting safety system scram settings are at values which will not allow the reactor to be operated above the safety limit during normal operation or during other plant operating situations which have been analyzed in detail (4,5,6,7).

In addition, control rod scrams are such that for normal operating transients the neutron flux transient is terminated before a significant increase in surface heat flux occurs. Scram times of each control rod are checked each refueling outage to assure the insertion times are adequate. Exceeding a neutron flux

. scram setting and a delay in the control rod action to reduce neutron flux to less than the scram setting within 0.95 seconds does not necessarily imply that fuel is damaged; however, for this specification a safety limit violation will be assumed anytime a neutron flux scram setting of the APRM's is exceeded for longer than 0.95 seconds.

Analysjs withlu the nominal uncertainty range of all appropriate significant parameters, show that if the scram occurs such that the neutron flux dwell time above the limiting safety system setting is less than 0.95 seconds, the safety limit will not be exceeded for normal turbine or generator trips, which are the most severe normal operating transients expected.

The computer provided with Monticello has a sequence annunciation program which will indicate the sequence in which scrams occur such as neutron flux, pressure, etc.

This program also indicates when the scram set point is cleared. This will provide informatisn on how long a scram condition exists and thus provide some measure of the energy added during a transient. Thus, computer information normally will be available for analyzing scrams; however. -if the computer information should not be available for any scram analysis, 1

Specification 2.1.C.2 will be relied on to determine if a safety limit has been violated.

During periods when the reactor is shut down, consideration mest also be given to water level requirements due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core will be cooled sufficiently to prevent clad melting 'should the water level be reduced to two-thirds the core height. Establishment of the safety limit at 12 inches above the top of the fuel provides adequate margin. This level will be continuously monitored whenever the recirculation pumps are not operating.

(4) FSAR Volume I, Section 111-2.2.3 (5) FSAR Volume III, Sections XIV-5 (6) Supplement on Transient Analyses submitted by NSP to the AEC February 13, 1973 (7) Letter from NSP to the AEC, " Planned Reactor Operation from 2,000 MRD/T to end of cycle 2", dated August 21, 1973 16 2.1 Bases REV pgp 373

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2.3 the vorst case IC1IFR during steady state operation is at 110% of rated power. Feaking factors as specified in Section 3.2 of the FSAR were considered. The total peaking i actor was 3.00.

The actual power distribution in the core is established by specified cont rol ral sequences an I is monitored continuously by the in-core LPFJ4 system.

As with the APIU1 scrar. setting, the IsPR 1 rod block setting is adjusted downward if peaking factors greater than 3.08 0:ist. This assures a rod block will occur. before 11CHFR beomes less than 1.0 even for this decra led ase.

The r 21 block setting is chanced by' changing the intercept point of the flow bias carve (keeping the slope. con 3 tant); thus, the entire curve will be shif ted downward.

The operator will set t he APRM rod block trip settires no greater than tu,t cham in Ficure 2. 5.1.

ilowever, the act u il set point can be as much as 3% creater t h ui th,t ch+ sen on Fi cu rt f.3.1

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circulation driving flows-less than 50% of design and 25 creater thm that chuwn Ibr rt e i ".n l :.. c l

driving flows Creater than 50% of design due to the deviations 'discusse i on r,ce 18.

C.

Reactor Low Wat er Level Scram - The reactor low water level..rran

.tY : point wh16h iil assure that the water levtl used in the bases fer the n,fety linit i s raintaine 1.

The operator will set the low water level trip setting no lower than lO'o" above thc top of inc active fuel. However, the actual set point can be as much as 6 inchen lower due to the deviations i

discussed on Page 18 D.

Reactor Iow Low Water Level ECCS Initiation Trip Point - The "mergency core coolin.; subsystems are designed to provide sufficient cooling to the core to dissipate the energy associats.d with the loss of coolant accident and to limit fuel clad temperature to well telow the clad r.elt inc temperature to assure that core Ceometry rec.ains intact and to limit any clad metal-water reaction to less than 1%. The desicn of the ECCS co:r.ponents to meet the above criterion was depenlent on three previously set parameters: ~ the maximum break size, the low uster level serm set point.

and the ECCS initiation set point. To lower the set point for initiation of the ECCS vull prevent the ECCS components from meeting their criterion. To raise the ECCS initiation set point would be in a safe direction, but it would reluce the margin estatliched to p event actuation of the ECCS during normal operation or durinC normally expecto1 transients.

2.3 BASES 21 Ert DCT 7 1973

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2.0 SAFETY LIMITS LIMITING SAFETY SYSTEM SETTIIES i

L 2.2 REACTOR COOIRff SYSTEM 2.4 REACIOR 000IRiT SYSTEM Applicability:

Applicability:

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Applies to limits on reactor coolant system Applies to trip settings of the instruments L

pressure.

and devices which are provided to prevent the l

l reactor system safety limits from being ex-(

l ceeded.

Objective:

Objective:

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To establish a limit below which the inte6rity To define' the level of the process var. tables 4

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of the reactor. coolant system is not threatened at which automatic protective action in

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due to an overpressure condition.

initiated to prevent the safety limits from i

being txceeded.

Specification:

Epeci fica t i,o_n:

The reactor vessel pressure shall not exceed A.

Reactor Coolant High pressure Scram shall

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1335 psig at any tLme when irradiated fuel is be $ 10/5 psig.

present in the reactor vessel

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B.

Reactor Coolant System Safety / Relief Valves l

Initiation shall be as follows:

2 h valves at f 1060 psig.

C.

Reactor Coolant System Safety Valves Hominal Settings shall be as follows:

h Valves at 21240 psig.

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_ Bases:

i 2.2 The reactor coolcat system integrity is an important barrier in the prevention of uncontrolled release of fission products.

It is essential that the integrity of this system be protected by i

establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuel in the reactor vessel.

The pressure safety limit of 1335 psig as measured in the vessel steam space is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The 1375 psig value was derived from the de.+1gn pressures of the reactor pressure vessel, coolant piping, and recirculation pump casing.

1 ie rs 'ective design pressures are 1250 psig at 575'F,1148 psig at 562*F, and 1400 psig at 575'F.

"The pressure safety limit was chosen as the lower of the pressure transients permitted by the applicable design codes: ASME Boiler and Pressure Vessel Code Section III-A for the pressure vessel,

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ASME Boiler and Pressure Vessel Code Section III-C for the recirculation pump casing, and the USAS Piping Code Section B31.1 for the reactor coolant system piping. The ASME Code permits pressure transients up to 10 percent over the vessel design pressure (110% x 1250 = 1375 psig) and the USAS Code permits pressure transients up to 20 percent over the piping design pressure (120% x 1148 =

1378 psig).

The design basis for the reactor pressure vessel makes evident the substantial margin of proteetion against failure at the safety pressure limit of 1375 psig. The vessel has been designed for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig and temper-ature of 575*F; this is more than a f actor of 1.5 below the y.'. eld strength of 42,300 psi at this tempera tu re ~ At the pressure limit of 1375 psig, the general membrane stress increases to 29,400 psi, still safely below the yield strength.

The reactor coolant system piping provides a comparable margin of protection at the established pressure safety limit.

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Bases Continued:

2.2

. Ibe nonaal operating pressure of the reactor coolant system is approximately 1025 psig. The turbine trip with failure of the bypass system represents the most aevere primary system pres-sure increase, resulting from an abnormal operational transient. The peak pressure in this transient is limited to 1214 psig. The safety valves are sized assuming no direct scram during IGIV c1csure. The only scram assumed is from an indirect means (high flux) and the pressure at the bottom of the vessel' is limited to 1306 psig in this case. Reactor pressure is continuously monitored.in the-control room during operation on a 1500 psig full scale pressure recorder.

2.2 Bases 25 REV 007 e 1973

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Bases:

2.4 The settings on the reactor high presst r e seram, reactor coolant system safety / relief valves, turbine control valve fast closure scran, and t u:bine stop valve closure scram have been established to assure never reaching the reactor coolant r.ystem pressure snfety limit as well as assurir.6 the sys-l tem pressure does not exceed the rance of the fuel cladding integrity safety limit. The AIBit neutron

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flux scram and the turbine bypass system also provide protection for these safety limits.

In addition to preventing power operation above 1075 psig, the pressure ceram backs up the AIPM neutron flux scram for steam line isolation type transients.

t The reactor coolant system safety valves of fer yet another protective feature for the reactor t

coolant system pressure safety limit.

la conpliance with Section III of the ASMC Boiler and Pressure vessel Code, 1965 editioa, the safety valves nust be set to open at a pressure t.o higher than 105 percent of design pressure, and they most limit the reactor pressure to no core than 110 percent I

of design pressure. The safety valves are sized according to the code for a corulition of !!SIV closure while operating at 1670 int, f-11 owed by no MSIV closure scram but scram f rom an indirect (high flux) means. With the safety valves set as specified herein, the maximum vessel pressure (at the bottom of the pressure vessel) would be about 1303 psig. Sec FSAR Section 4.4.3 and i

supplemental information submitted Febr uary 13, 1973.

Evaluatious presented indicate that a total of eight valves (h safety valves and h dual purpose safety / relief valves) set at the specified pressures matutain the peak pressure during the transient within the code of allowable and safety limit pressere.

The operator will get the reactor coolant high pressure scram trip setting at 1075 psig or lower.

l However, the actual setpoint can be as much as 10 psi above t he 1075 psig indicate 1 set point due l

to the deviations discussed in the basis of Specification 2 3 on Page 18.

In a like nanner, the operator will set the reactor coolant system safety / relief valve initiation trip setting at 1080 psig or loser. However, the actual set poin' can be as much as 11 psi above the 1030 psig indicated l

set point due to the deviations discussed in, e basis of Epecification 2 3 on Page 18.

A violation of this specification is ass'uned to occar only when a device is knowingly set outside of the limiting trip setting, or when a sufficient number of devices have been affected by any means 2.4 EASES 26 REV gg7 p 373 4

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2 Bases Continued 3.3 and 4.3:

I consequences of reactivity accidents are functiona of the initial neutron flux. The require-j ment of at Ieast 3 counts per second assures that any trant.ient, should it occur, begins at

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or above the initial value of 10% of rated power used in the analyses of transients from cold

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i' conditions. One operable SRM channel would be adequate to monitor the approach to criticality i

using homogeneous patterns of scattered control rod withdrawal. A minimum of two operable

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SRM's are provided as an added conservatism.

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5.

The consequences of a rod block monitor failure have been evaluated and reported in the Dresden 11 SAR Amendments 17 and 19.

Thesa evaluations, equally applicable to Monticello, show that i

during _ reactor operation with certain limiting control rod patterns, the withdrawal of a designated single control rod could result in one or more fuel rods with MCHFR's less than 1.0.

During use of such patterns, it is judged that testing of the RBM system prior to withdrawal of such rods to assure its operability will assure that imprcper withdrawal does not occur. It is the responsibility of the Engineer, Nuclear, to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable rods in other than limiting patterns.

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Scram Insertion Times l

The contml rod system is designed to brind the reactor suberitical at a rate fast enough to prevent I

fuel dnmnge; i.e., to prevent the MCHFR fmm becoming less than 1.0.

'Ihis requires the negative l

reactivity insertion in any local region of the core and in the over-all core to be equivalent to at least one dollar within 0.75 second. 'Ibe req'timd average scram times for three control rods in all two by two arrays and tl.e required average scram times for all control rods are based on inserting this amount of negative reactivity locally and in the overall core, respectively, within O.75 second.

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t Under these conditions, the ther=al limits are never reached during the transients requiring control j

rod scram as presented in the FSAR.

'Ibe limiting operational transient is that resulting fram a turbine f

i stop valve closure with fai. lure or the turbine bypass system. Analysis of thic transient shows that the l

negative reactivity rates resulting from the seram with the average response of all the drives us given j

in the above Specification, provide the required protection, and MCliFR remains greater than 1.8.

In the analytical, treatment of the transients, 290 milliseconds are allowed between a neutron sensor j

reaching the scram point and the start of motion of the control rods.

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3 3/h.3 BASE 3 85 l

l W CCT 2 1973 l.

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30 LIMITING CONDITIONS FOR OPEMTION h.0 SURVLILLANCE BFQUIREMEITfS

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(b) When the ecntinuous conductivity meni-tar is inoperable, a reactor coolant sample should be taken at least once per shift and analyzed for conductiv-ity and chloride ion content.

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If Specific ation 3 6.C.1, 3 6.C.2, and 3 6.

l

. C.3 are not met, normal orderly shutdown i

shall be initiated.

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I D.

Ccolant Irakage D.

Coolant Leakage r

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j Any time irradiated fuel is in the reactor vessel, Pencter coolant system leakage into the dry-j and reactor coo:_ ant terIerature is above 2120F, well shall le checked and reccrded at leact t

j reactor coolant leakage into the p?icary contain-once per day.

j ment from unidentified sources shall not exceed i

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5 gpm.

In addit ion, the total reactor coolant I

system leakage into the primary aontainment shall l

not exceed 25 spm. If these conditions cannot be j

met, initiate ar orderly shutdown and have the re-l actor placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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E.

Safety and Beller Valves E.

Safety and Eelief Valves 1

1.

During power (perating conditions and whenever 1.

A minimum of two safety valves shall be 1

the reactor ecolant pressure is greater than bench checked or replaced with a bench 110 psig and temperature greater than 345 F, checked valve each refuelin6 outage. All 3

l four safety valves and the safety valve func-four valves shall be checked or replaced

'3.6]h.6 118 1

BEv OCT 2 573

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30 LIMITING CONDITIONS FOR OPERATICU 4.0 SURVEILLANCE FEQUIRDIEIffS t

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l tion of four safety / relief valves shall every two refueling outages. The nominal

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be operable. The solenoid activated popping point of the four safety valves i

relief function of the safety / relief val-shall be set at s 1240 psig.

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I ves shall be operable as required by Spec-ification 3 5.E.

4 2.

If specificatica 3 6.E.1 is not met, ini-

2. a. A uinimum of two safety / relief valves shall tiate an orderly shutdown and have coolant be tench checked or replsced with a tench j

pressure and temperature reduced to 110 checked valve each, refueling cutage.

All psig or less and 3kS F cr less within 2h four valves shall be checked or replaced I

O hours.

every two refueling outages. The popping t

point of the safety / relief valves stall be set as follows-l Number of Valves Set Point (psig) 4

< 1080

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b.

At least one of the safety / relief valves i

shall be disassembled and inspected each refueling cutage.

c.

The integrity of the safety relief valve bellows shall be continuously monitored.

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d.

The operability of the bellows monitoring l

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l 36/4.6 t

REV OCT g 1973 i

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Bases Continued 3.6 and 4.6:

D.

Ccolant Leakage j-The former 15 gpm limit for leaks from unidentified sources was established assuming such leakage was coming I

from the primary aystem. Tests have been conducted which demonstrate that a relationship exists between the size cf a crack and the probability that the crack will propagate. From the crack size a leakage rate can be determined.

For a crack size which gives a leakage of 5 gpm, the probability of rapid propagation is less than 10-5 hus, an j

unidentified leak of 5 gpm when assumed to be from the primary system had less than one chance in 100,000 of propa-j gating, which provides adequate margin. A leakage of 5 gpm is detectable and measurable. De 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period allowed for determination of Icakage is also based on the low probability of the crack propagating.

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ne capacity of the drywell sump pumps is 100 gpm and the capacity of the drywell equipment drain tank pumps is also 100 gpm.

Removal of 25 gpm from either of these sumps can be accomplished with considerable margin.

i ne performance of the reactor coolant leakage detection system, including an evaluation of the speed and sensi-tivity of detection, will be evaluated during the first 18 months of plant operating, and the conclusions of this evaluation will be reported to the AEC. Modifications, if required, will be performed during the first refueling outage after AEC review.

In addition, other techniques for detecting leaks and the applicability of these techniques ro the Monticello Plant will be the subject of continued study.

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E.

Safety and Relief Valves Experience in safety valve operation shows that a testing of 507. of the safety valves per refueling outage is i

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adequate to detect failures or deterioration. A tolerance value is specified in Section III of the ASME Boiler and l Pressure Vessel Code as +17. of the set pressure. An analysis has been performed which shows that with all safety valves set 17. higher than the set pressure, the reactor coolant pressure safety limit of 1375 psig is not exceeded.

L 9afety/ relief valves are used to minimize activation of the safety valves. The operator will set the pressure settings at or below the settings listed. However, the actual setpoints can vary as listed in the basis of Specification 2.4.

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he required safety valve steam flow capacity is determined by analyzing the pressure rise accompanying the main

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4 steam flow stoppage resulting from a MSIV closure with the reactor at 1670 MWt.

The analysis assumes no MSIV i

closure scram, but a reactor scram from indirect means (high flux). We relief and safety valve capacity is

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assumed to total 83.97. (477. relief and 36.97. safety) of the fall power steam generator rate.

his capacity I

3 corresponds to assuming that four safety / relief valves (477.) and four safety valves (36.97.) operated.

I 3.6/4.6 BASES 134 i

j REV DCI 2 1973 t

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