ML20024G612

Application for Amend to License DPR-22,implementing Mods Described in Permanent Plant Changes to Accommodate Equilibrium Core Scram Reactivity Insertion Characteristics.

ML20024G612
Person / Time
Site:	Monticello
Issue date:	03/01/1974
From:	Larkin W NORTHERN STATES POWER CO.
To:
Shared Package
ML20024G609	List: ML20024G608 ML20024G612 ML20024G613
References
NUDOCS 9102130482
Download: ML20024G612 (14)

Text

_ _ . - . - - . . . - _ _ _ _ _ - _ _ _ _ . __ - - . _ - - - . - - - _ - - - - __,__-. . . -

1 lmITED STATES AIDMIC ENERGY C0!D:ISSION NORTilERN STATES POWER COMPANY Monticello Nuclear Generating Plant Docket No. 50 263 REQUEST FOR AUTt!0RIZATION OF A CllANGE IN TECtlNICAL SPECIFICATIONS Or APPENDIX A PROVISIONAL OPERATING LICENSE NO. DPR-22 (Change Request Dated March 1,1974)

Northern States Power Company, a Minnesota corporation, requests authorization for changes to the Technical Specifications as shown on the attachments labeled Exhibit A and Exhit't B. Exhibit A describes the proposed changes along with reasons for change. Exhibit B is a set of Technical Specification pages incorporating the proposed changes.

This request contains no restricted or other defense information.

NORTilERN STATES POWER COMPANY By i

//' .

Wade Larkin Croup Vice President - Power Supply 4, ; /

On this I day of 8 Nov ,[d@before nie a notary public in and for said County, personally appeared Wade Larkin, Group Vice Presi-dent - Po'rer Supply, and being first duly sworn acknowledged that he is authorized to execute this document in behalf of Northern States Power l Company, that he knows the contents thereof and that to the best of his knowledge, information and belief, the statements made in it are true and that it is not interposed for delay.

,s }

f . ,.- f sjj ,

)-y- ?,m/ k 'khlIS

/' ; John J Spith l b-JOHN J. SMITH Notary Puble Hennenin Ccunty, M ena'nta My Comm:ssion Espaes Mmh 3,1976 l

l 9102130482 740301 PDR ADOCK 05000263 P PDR

,____u

_ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ , . _ m _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ .

i. ,

1

) EXHIBIT A

, MONTICELLO HUCLEAR GENERATING PIANT l

4 DOCKET NO. 50-263 CHANGE REQUEST DATED - MARCH 1,1974 PROPOSED CHANGES TO TECIDiICAL SPECIPICATIONS APPENDIX A OP PROVISIONAL OPERATING I f

LICENSE NO. DPR-22 j l

1. PROPOSED CRANCE l

j On Page 22, Bases 2.3.E. , insert "and PRT" af ter ". . .f ast closure scram."

in the first sentence of paragraph 2.3.E. ,

REASON FOR CHANGE This shows that the PRI system also assists in limiting the transient effects ,

i of control valve fast closure.

l 2. PROPOSED CHANGE i -

On Page 22, Bases 2.3.E, revise the last sentence of paragraph 2.3.E to read: '

" Reference PSAR sections 14.5.1 and 14.5.2 and supplemenral information

entitled " Permanent Plant Changes to Accommodate Equilibrium Core Scram

Reactivity Insertion Characteristics" submitted January 23, 1974 I

1 REASO': FOR CHANGE

, The analytical justification for the proposed changes is included,in the

PRT submittal and is appropriately referenccd here.

3. PROPOSED CRANCE On Page 22, Bases 2.3.P. revise the paragraph to read as follows:

"Ihe turbine stop valve scram with PRT, like the load rejection J scram with PRT, anticipates the pressure, neutron flux and heat flux increase caused by the rapid closure of the turbine stop i

valves and failure of the bypass. With a setting at 10% of valve i closure for scram, and PRT, the increase in heat flux is limited such that adequate pressure and thermal margins are maintained.

The PRT opens safety / relief valves to limit the pressure and heat flux increases, and allows safety / relief valve reclosure as pressure 4

decreases. Por this event, the peak surface heat flux and MCHPR remain within limits. Reference PSAR Section 14.5.1.2.2 and supple-mental information submitted January 23, 1974."

,, - ,a ..~-- w-

- - _ ~ - - - - - - - - - - . _ -

l s ,

e ,

e s

REASON FOR CllANCE The insertion of the PRT discussion reflects the changes of the transient event as a result of the PRT addition. We elimination of specific heat ,

flux and MCilFR numbers will aid in minimizing future tech spec changes  :

should these numbers change. Replacement of the February 13, 1973 reference with the January 23, 1974 submittal "pdates the bases to the latest analysis.

4. PROPOSED C11ANCE On Page 23. TS 2.4.B. revise to read as follows:

"B. Reactor Coolant System Safety / Relief Valves shall be set l as follows:

6 valves at 41080 psig."

f REASON FOR CilANCE i

1 This specification is changed to reflect the required number of safety /reiicf 1 valves based on the MSIV closure event with indirect scram (ASME overpresdure  ;

test).

l S. PROPOSED C11ANCE

! On Page 23, TS 2.4.C. delete the existing specification.

I RFASON FOR CilANCE i

! This change climinates reference to safety valves.

i 6. 1ROPOSED CHANGE i i On Page 25, TS Bases 2.2, revise the end of the third sentence from '

! "..., is limited to 1214 psig " to ".... is limited to 1178 psig at

! the bottom of the vessel." Insert in place of the fourth, fif th, and l sixth sentences; "The primary system overpressure protection analysis assumes the closure of all MSIVs with indirect (high flux) s(cam. Peak

! pressure at the vessel bottom is 1285 psig."

i L f REASON TOR C11ANCE i

! nis chante reflects the results of the current analyt.is. (Supplemental  !

l information submitted January 13, 1974.) In addition, reference to  :

l continuous monitoring of reactor pressure on a 0-1500 psi recorder is deleted.

l This eliminates an error in the bases since only a 0-1200 psi recorder is l available for continuous service. Transient pressures above 1200 psi are ,

measured on a 0-1500 psi gauge which marks peak pressure.

l l

l

- 2-I

1 O

t

7. PROPOSED CHANGC Dn Page 26, TS Basea 2.4, revise the first paragraph, first sentence to include " prompt relief trip system," following "high pressure scram,",

Revise the second paragraph as follows:

Lines 1,3,5 6 7 Change " safety" to " safety / relief" Line 8 Change "1308" to "1285" Line 9 Change "rebruary 13, 1973" to " January 23, 1974" Line 10 Change ta read as follows:

". . .. of six dual purpose safety / relief valves set at 1080 psig maintain the peak pressure during the event within the limits allowed by the ASME Code."

Add the following sentence to the third paragraph:

" Analyses were performed asseming a safety / relief valve setpoint of 1080 psig + 1%."

REASCG FOR CRANCE These changes reflect the results of the current analyses.

8. PROPOSED CHANGE On Page 39. TS Bases 3.1, first paragraph, the third and fourth sentence =

are changed to read as follows:

"The turbine stop valve closure scram with PRT adequately preserves the margins to pressure and MCHFR limits should a turbine trip with bypasc failure occur."

Revise the fifth sentence of the first paragraph and the third sentence of the third paragraph; change "rebruary 13, 1973" to " January 23, 1974."

REASON FOL CHANGE These changes recognize the addition of the PRT system.

9. PROPOSED CHANGE on page 49, add a new paragraph 3.2.F as follows:

"F. Prompt Relief Trip (PRT) System The limiting condition for operation for the instrumentation that initiates the prompt relief trip are given in Table 3.2.4.A.

REASON FOR CRM9CE A separate new specification is required to cover the PRT system.

e

10. PROPOSED CllANGE On New Pages 60A and 60B add Table 3.2.4A (See Appendi B of this change request.)

RIASON POR CllAliCE

'Ihe PRT System Instruments require a separate table to define operability to insure differentiation f rom other protective functions.

11. PROPOSED citANGE On Page 62, TS Table 4.2.1, insert the following af ter "orr CAS ISOLATIm."

(INSTRUMD;T CilANNEL) (TEST) (CALIBRATION) (SC;SOR CllECK)

PRmPT RELIEr TRIP (PRT)

SYSTIE

1. PRI Disable Note 1 Onca/3 months None (1:cactor Lov Pressure) 2 PRT Power Range Note 1 Once/3 months None Permissive (Turbine First Stage Pressure)

3. PRT Timer Note 1 Once/3 months None 4 Turbine Stop Valve Note 7 None None Closure

5. Turbine Control Valve Note 7 None None Closure REASO.; ron CilANGE Test and calibration frequencies for PRT instruments not included elsewhere must be provided.

12. PROPDSED CilANGE On Page 63, TS Table 4.2.1, add the following note:

"(7) Instrument functional test and calibration shall include ver:.fication of instrument channel response in the PRT sy s t em. Frequencies are established in TS Tables 4.1.1 6 4.1.2."

PIASON FOR, CllMQE

'1hi6 provides specific functional requirements for subject Reactor Protection System instrumen ts relative to PRT.

1 9

q 1

i i I 13. PROPOSED CHANGE f

j On Page 64,1S Bases 3.2, add the words "and the Prompt Relief Trip System." to the last sentence of the first paragraph.

REASON FOR CHANGE 1

The PRT system is included in the TS section on Protective Instrumentation,

] The above referenced paragraph indicates that all systems in this section j

! except contral rod block is single failure proof even during testing. The PRT system thould be listed with control rod block in thi,s aspect.

I

! 14 . PROPOSED CHANCE on Page 68, TS Bases 3.2 add the following new paragraph after the i existing third paragraph, continuing on a new Page 68A, and move the existing fourth paragraph on Page 68 to new Page 68A.

"The prompt relief trip (PRT) system initiates the opening of t

three or six safety / relief valves at reactor power levels 270%

4 and 2 85%, respectively, with the occurrence of a turbine stop' l valve or control valvo closure. The PRT initiating action originates i in two independent channels, each capabic of satisfying systen j requirements through redundant instruments. Interval timers with a low pressure back-up disable the PRT; self actuated pressure operation of the safety / relief valves remains unaffected. The settings of the j instruments provided in Table 3.2.4A ensure that pressure and thermal margins are maintained during the worst-case single-failure-caused abnormal operational transient, i.e., turbine trip with failure of the bypass valves. In addition, the PRT utilizes a 1 out of 2 logic system. In accordance with IEEE-279 an exception is taken to the minimum operating requirements to allow a short period of time, during which an instrument channel may be bypassed to allow for testing. For this logic the single failure protection is temporarily defeated."

1 j REASOS FOR CHANGE A bases statement that generally describes the PRT is required for nev

TS 3.2.r. This statement follows the format of others in this section.

s 5-

. - - . . , --w, - . - , , - - , , - , ._-.-y....m.---....--y. -

t *m7 -'*

15 . PROPOSED CitANGE i

on Page 70, TS Table 3.2.5, insert the following new entry in the Table:

5

}~

Instrumeritation that TRIP FUNCTI(N DEVIATION Controls the prompt relief 1 trip (PRT) System PRT Disable i 1%

j (Reactor low pressure) 1 PRT Timer 2 0.5 sec.

j PRT Power i 3% of rated Range Permissive pressure l

(Turbine First j Stage Pressure)

I REASON FOR CllANT

, Deviations must be presented to account for variances in settings given in

] Table 3.2.4A. These numbers correspond to the variability allowed in the

analyses.

L

)

16. PROPOSED CHANGE On Page 79, TS 3.3.1.0, in TS 3.3.1.C.1 change the 90% insertion time from "S.00" to "3.50" and in TS 3.3.1.C.2 change the 90% insertion time from "5.30" to "3.80."

REASON POR CHANGE

These changes reflect the "67B" rod scram cimes that better describe control i rod performance. The 67B times were used in all analyses performed in support of the reload 2 and " permanent plant changes."

17 . PROPOSED CilANGE i On Page 85, TS Bases 3.3.C, revise the first and second sentences to read as follows:

"The control rod system is designed to bring the reactor suberitical at a rate fast enough to ensure the maintenance of edequate fuel thermal and reactor pressure margins for non-c cident events. This requires the negative reactivity insertion in any local region oi the core and in the overall core to be at 1 cast as great as the (End-of Cycle equilibrium core) scram reactivity insertion curve used in the analyses '

submitted on January 23, 1974."

Delete "within 0.75 seconds" from the third sentence.

' Add "and the supplemental information submitted January 23, 1974" to the end of the fourth sentence.

i In the sixth sentence change "1.8" to 1.35."

I .

6-

. _ - _ _ _ - , - _ _ . - _ ~ .-- _ _ - _ __-

REASON FOR CHANGE These changes correct the bases to reficct the relationship between the rod scram times and the scram reactivity curve, and serve to incorporate the results of current analyses into the bases. A $/sec number was originally defined by the scram reactivity curve (rather than the converse); to prevent coniusion, the bases statement has been changed.

18. PROPOSED CHANGE On Page 112, TS Bases 3.5.E, change the beginning of the fourth sentence te read "Three safety / relief valves are included . . . ."

RTASDN FOP CHANGE With the addition of safety / relief valves the number given for the totc1 is not correct. The listing of the total number of safety / relief valves is not applicable to the bases statement.

19. PROPOSED CHANGE on Page 115, TS 4.6. A.1, revise to read as follows:

"1. During heatups and cooldowns recirculation loops A and B temperatures shall be permanently recorded at 15 minute in t e rvals . "

REASON FOR CHANGE Deletion of the requirement to establish permanent records of reactor vessel shell and shell flange temperatures is warranted since outside vessel wall temperatures are not used to demonstrate compliance with Technical Specification 3.6. A Themal Limitations. 'Ihe purpose of the thermocouples located on the outside surface of the reantor vessel is to provide a means of observing vessel temperature in response to vessel coolant temperature changes. The behavioral patterns observed should generally follow those of the coolant within the associated vessel region. Temperatures will generally lag coolant temperatures and transient effects will be masked. , The knowledge gained from the reactor vessel surface thermocouples on operating reactors has been reflected in the coolant temperature plots shown on current reactor vessel thermal cycle diagrams (in particular, the occurence of a cold water layer in the bottom head region of the reactor vessel due to drive water inleakage under low reactor flow conditions was revealed by vessel surface thermocouples). However, there is no reason to continuously monitor vessel surface temperatures or temperature differentials beyond the period of startup and power operation testing. Measurements and recordings of the temperature and flow condition of all coolant systems entering and leaving the reactor vessel together with vessel pressure and water icvel recordings should provide an adequate index of reactor vessel thermal conditions during subsequent plant operacions since these are key operating parameters which are used to define thermal stresses. Reactor saturation temperature (obtained by pressure-temperature conversion) should be used to define vessel upper region temperature. Reactor recirculating loop and vessel bottom drain temperatures, should be used to define zaactor vessel downcomer and lower region temperatures, respectively.

J

13 . Pkop0 SED CilANGE on Pages 118 and 119, TS 3.6.E, delete the existing specification entirely and replace with the following.

"E. Safety / Relief Valves and Prompt Relief Trip (PRI) System

1. During power operating conditions and whenever reactor coolant pressure is greater than 110 psig and temperature is greater than 345 F;

a. The safety valve function (self-actuation) of six

~

safety / relief valves shall be operable,

b. The solenoid activated relief function (Automatic Pressure Relief) shall be operable as required by Specification 3.5.E.

2. During reactor power operation, the prompt relief trip -

(PRT) system function of six safety / relief valves shall be operable in accordance with Specification 3.2.F."

RFASON FOR CllANGE The requirements for Automatic iressure Relief and safety / relief velve self-actuation remain the same except that six self-actuated safety / relief valves rather than four are required as described in the January 23, 1974 submittal. The operational limitations on these factors remain unchanged.

New specification 3.6.E.2 establishes *' operability requirement for valves on the PRT System. Specific lim. -

conditions are established in Table 3.2.4A.

20. PROPOSED CHANGE G. Pages 118, 119, and 120, TS 4.6.E, delete the existing specification entirely and replace with the following.

"t, bafety/ Relief Yalves and Prompt helief Trip (PRI) System

1. a. A minimum of six safety / relief valves shall be bench checked or replaced with a bench checked valve each refueling outage.

The nominal setpoint of all operational safety / relief valves shall be c1080 psig,

b. At Icast two of the safety / relief valves shall be disassembled and inspected each refueling outage,

c. The integrity of the safety / relief valve bellows shall be continously monitored.

d. The operability of the bellows monitoring system shall be demonstrated at least once every three months.

.8 .

1

_ _ - _ _ _ _ _ _ _ _ . _ _ _ _ . . -- _ _ _ _ - _ _ - - - - - - - _ - _ _ _ _ - - - - - - - - - - _ - - - - - _ - - - - - - - - - - - - - - ~

I

2. Surveillance of the PRT System shall be as follows:

Iten Frequency Valve Operability Each Operating Cycle Simulated Automatic Actuation Test Each operating Cycle RI FOR CRANGE

.nges provide the surveillance requirements for the safety / relief s ased on the analyses of the Jan. 23, 1974 report and availability requirements based on ASME codes.

The operability tests are generally of the same type and frequency as systems in rervice of comparable importance.

20. PROPOSED CHANGE On Page 134, TS Bases 3.6.E and 4.6 E, replace the last two paragraphs on the page with the following:

E. Safety / Relief Valves and Prompt Relief Trip Testing of all safety / relief valves each refueling outage ensures that any valve deterioration is detected. A toler-ance value of 1% for safety / relief valve setpoints is specified in Section III of the ASME Boiler and Pressure Vessel Code.

Analjses have been performed with all valves assumed set 1%

higher (1080 psig + 1%) than the nominal setpoint; the 1375 psig code limit is not exceeded in any case.

The safety / relief valves are used to limit reactor vessel overpressure and fuel thermal duty through prompt relief trip and self actuation.

The required safety / relief valve steam flow capacity is determined by analyzing the transient accompanying the main steam flow stoppage resulting fro. a postulated MSIV Closure from a power of 1670 MWt .

The analysis assumes a multiple-failure wherein direct scram (valve position) is neglected. Scram is assumed te be from indirect means (high flux).

In this event, the safety / relief valve capacity is acsuacd to be 71%

of the full power steam generation rate."

REASON FOR CRANCE Changes to this section are necessary to explain the new safety / relief valve arrangement and code basis for analyses and testing of the valveo. Wording and numbers have been changed accordingly.

21. PROPOSCD CHANGE on Page 135, TS Bases 3.6 and 4.6 E, replace the first three paragraphs on this page with the following:

"The safety / relief valves have two functions; i.e. automatic vessel depressurization or over-pressure protection. The former is a solenoid actuated function (Automatic Pressure Relief) in which 1

external instrumentation signals of coincident high drywell pressure and low-low water Icvel initiate opening of the valves. This function provides backup to the HPCI system for small break protection and is discussed in Specification 3.5.E. In addition these valves can be operated manually via an independent solenoid.

The over-pressure protection function utilizes six safety / relief valves, three of which are operated for the Automatic Pressure Relief function. All six valves are capable of direct, self-actuation or indirect, prompt-relief trip (PRT) actuation.

The primary overpressure pro,tection (for ASME Code consideration) is provided via the pressure-actuated integral bellows and pilot valve that cause main valve operation for any plant event therein valve setpoint pressure is attained. Article 9 of the ASME Pressure Vessel Code Section III, Nuclear Vessels, requires that the bellows be monitored for failure since this wculd defeat the function of the safety / relief valve."

REASO': TOL CHANGE Safety / relief valve functions and Code considerations are discussed here. These changes accommodate the new arrangement.

23. PROPOSED CHANGE On Page 136, TS Bases 3.6 ar.d 4.6.E. , move the remainder of the fif th paragraph back from page 137. >bve the first full paragraph from page 137 to the top of new page 136?.. Then insert the following on new page 136A, finishing at the top of page 137.

"The prompt relief trip (PRI) function of the safety / relief valves provides an anticipatory actuation of the safety / relief valves for tranaients involving turbine stop valve closure or turbine control valve fast closure. Although the PRT system is intended to provide anticipatory pressure relief for turbine trip transients with failure of the bypass valves, the PRT system is entirely independent of the bypass system. The IRT system, by providing an anticipatory open signal to the safecy/ relief valves aids in maintaining fuel thermal and pressure margine and has t.herefore been designed on the basis of Engineered Safeguards and meets the requirements of IEEE 279.

The PRT system 4.s divided into two independent redundant channels which are programmed on a power icvel schedule into three modes of operation relative to the safety / relief valves coupled to it. The modes of operation are listed below:

Reactor Power PRT/S/RV's

} 857, 6

) 707, 3

< 70% 0 1

k Mode selection is automatic through a biasing signal based on tur'

• ae first stage pressore. The power level schedule has been esta ' Lshed to ensure the full power transients remain the most limiting and adequate pressure and fuel thermal margins are provided.

The PRT system will not preclude safety / relief valve self actuation or Automatic Pressure Relief Operation.

Deactivation of the PRT system signal is ef fected through redundant interval thmers and a 1sw reactor pressure switch or the attainment c f the scif-actuation reset pressure, should the pressure rem..n above the self-actuation setpoint for a period exceeding the interval thmer setting.

Spare safety / relief valves may be installed that are adaptable to PRT service with minor interco..nection changes to permit the substitution for an inoperable PRT valve while operating in a reduced power mode.

The PRT Syste.n is incore ' rated to t-ffset changes in the scram reactivity insertion rate which c.. .r with increasing exposure out ts the equili-brium exposure. All transients have been analyzed to account for botn L

, the slower insertion rates and PRT installation in " Plant Changes to Accommodate Equilibrium Core Scram Reactivity Irsertion Characteristics".

)

This report was submitted to the Commission in January 23, 1974.

The PRT systen demonstrates substantial improvement in fuel thermal and pressure respense to the PRT-coupled turbine and generator trip transients.

This effect is manifested in the results of a turbina trip with bypaso failure transient where PRT reduces vessel peak pressure by 74 psi and heat flux by 137. compared with the same event without PRT. PRT provides effective compensation for effects on plant performance caused by the changing scram reactivity. The PRT 3-mode design provices flexibility in maintaining pressure and fuel thermal margins and minimizing the duty cycle on the safety / relief talves at low power. Positive disabling P

of the PRT valve ection is ine.ured by a timer with a nominal setting of 5 seconds ard/or a low versel pressure signal cet at 950-970 psig.

RE250S FOR CHA!;GE This section provides a discussion of the PRT system from the standpoint of LCO's and Surveillance. As such, system operation in described and the reasons for the surveillance shown.

B b ,,. _ _ - _ _ _ , , _ _ _ - , _ _ _ _ _ _ _ _ _ _ _ , .-x _ _ _ , _ - _ _ - , _ _ _ _ _

24 PROPOSED CHANGES On Page 148, TS 4.7. B. add a "1." ahead of the sentence " Standby gas trectment system surveillanen shall . . . ."

On Page 148, TS 4.7.B.1.b, delete the first sentence through the colon and replace it with "During each refueling outage prior to refueling, whenever a filter is changed, whenever work is performed that could affect filter systems efficiency, and at intervals not to exceed six months between refueling outages, it shall be demonstrated that:"

On Pages 148 and 149, TS 4.7.B.1.b(1), delete the current sentence and replace it with "The removal ef ficiency of the installed particulate filters for particles having a mean diameter of 0.7 microns shall be equal to or greater than 997. based on an in-place dioctyi phthalate (DOP) test."

On Page 149, TS 4.7.B.1.c, insert the word "and between "shall be removed" and " adsorption".

On Pago 166, Bases 4.7, at the end of the first paragraph after "....

testing medium.", add a sentence as follows: " Individual filter units will be tested and certified to have a removal efficiency of equal to or greater than 99% for particles having a mean diameter of 0.3 microns at the time of purchase."

R.3ASON FOR CRANCE The addition of the "1" and the insertion of the word "and" are to correct omissior.c in the current Technical Specifications.

As presently written, Technical Specification 4.7.B.1.b requires ffiter tests (DOP 6 Freon) during each scheduled secondary containment leak rate test. It has been interpreted that this means filter testing must be simultaneous with lenk rate testing. This testing requirement is to demonstrate removal efficiency prior to refueling operations and is irrelevent to simultaneous tenti g with the leak rate test. The testing requirement is satisfied with tha proposed wording.

The document ORNL-NSIC-65, January,1970, " Design, Construction, and Testing of High-Efficiency Air Filtration Systens 10r Nuclear Applications", states that in-place tests of HEPA filters are made with a polydispersed aerosol of dioctyl phthalate (DOP) droplets having a light-scattering mean diameter of 0.7 p as opposed to quality assurrance tests of these filters, which are made with a ernodispersed DOP aerosol having a mean particle size of 0.3 u. AEC Regulatory Operations Inspection Report No. 050-253/73-12 for the Menticello Plant (Paragraph 10b) states in part that although currer.c Technical Specifications call for particles " larger than 0.3 micron", equipment which will generate particles of 0.9 micron size are not carrently available for field use. The inspection report in the same paragraph states tbat Licensing personnci disclosed that Standardized Technical Specf fications current 4r being developed for all facilities will clarify this matter and call for filter te. a u ing the 0.7 micron " cold" DOP test. The proposed change ir TS 4.7.B.1.b(1) and in the 4.7 Bases are required to clarify this point in the Monticello Technical Specifications.

l 1

-- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _. . _.J

EXHIBIT B This exhibit consists of the following pages revised to incorporate the proposed changes:

Page 22 Page 85 Page 23 Page 112 Page 25 Page 115 Page 26 Page 118 Page 39 Page 119 Page 49 Page 120 Page 60A Page 134 Page 60B -

Page 135 Page 62 Page 136 Page 63 Page 136A Page 64 Page 137 ,,

Page 68 Page 148 Page 68A Page 149 Page 70 Page 166 Page 79

- _ _ _ - _ _ - - - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ - - _ _ _ _ __ _ _