Text

Proposed Tech Specs as Shown on Attachments Labelled Exhibit a & Exhibit B
	ML20125C513
Person / Time
Site:	Monticello
Issue date:	07/01/1975
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML20125C501	List: ML20125C499; ML20125C503; ML20125C513;
References
	NUDOCS 9212110219
	Download: ML20125C513 (47)
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q DOIIBIT A MONTICELID NUCLEAR GENERATING PIANT DOCKE'I NO. 50-263 LICENSE AMENDMEITI REQUEST DATED JULY 1,1975 PROPOSED CHANGIE TO TECHNICAL SPECIFICATIONS APPENDIX A, OF PROVISIONAL OPERATING LICENSE DPR-22 Pursuant to 10CFR50 59, the holders of Provisional Operating License DPR-22 hereby propose the following changes to Appendix A, Technical Cpecifications:

1. SPECIFICATIONS 3/h.6. A AND 3/h.6.B, REACTOR VESSEL FRAC'IURE T0JGHNESS PROP 0GED CHANGE Replace pages 115,116,122,130, and 131 with the. corresponding pages fmm Exhibit B. Insert new pages 116A,122A,122B,1220, and 131A from Exhibit B.

REASON FOR CHANGE The Geneml Electric Company has perfomed a standant backfit analysis of the Monticello reactor vessel to determine proposed opemting limits based on 10CFR50 Apperdix 0. The results of this analysis have been incorporated in revised Epecifications 3/4.6. A and 3/4.6.B and Figures 3 6.1 through 3 6.h. When this revision is incorporated in the Monticello Technical Specifications, plant operating procedures and surveillance vill confom to 10CFR50 Appendix 0 and Appendix H.

2. PROP 0GED SPECIFICATION 3/h ll, SEALED SOURCE ColffAMINATION PROPOSED CHANGE Add new Specification 3/4.11, " Sealed Source Contamination," as contained in the Exhibit B pages 189B,1890,189D, and 189E.

RF160N FOR CHANGE Diis proposed chan6e is being submtted at the request of the Regulatory Staff. It provides for the leakage testing, inventory, stomge, and disposal of sealed radioactive sources.

9212110219 750701 PDR ADOCK 05000263 P PDR

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4 IXHIBIT A '

3. SPECIFICATION 3.2.D. AIR FJEWOR OFF-GAS SYSTIN PROPOSED CHANGE

a. Revise Specification 3 2.D.1 to read:

1. Except as specified in 3 2.D.2 and 3 2.D.3, both steam jet air ejector monitors shall be opernble during ,

reactor power operation. the trip settings for the air ejector monitors, except as specified in 3 2.D.h, shall te set to close the recombiner train inlet valve (s) within 30 minutes at a 3 radiation level not to exceed the equivalent of the maximum permitted ~

stack release rate after a decay time of 120 minutes.

b. Revise Specification 3 2.D.4 to read:

4. If operation is necessary with the Off-Gas Holdup System recombiners bypassed, the trip settin6s for the air ejector monitors shall be reset to close the stack off-gas isolation valve within 15 minutes at a radiation level not to exceed the equivalent of the maximum permitted stack release rate after a decay time of 30 minutes.

REASON FOR CHANGE The Technical Specifications now require the air ejector monitor trip setting to be less than the equivalent of the maximum permitted stack release rate based on a 30-minute decay period. The trip settings are now the same for all modes of operation of the Off-Gas System.

There are three principal modes in which the Off-Gas System may function, depending on the operability of various components in the system. They

< are:

bbde Estimated Holdup Time 1- Compressed Storage 50 - 250 hours0.00289 days 0.0694 hours 4.133598e-4 weeks 9.5125e-5 months 2- Recombiners Only 2 - 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months 3 - Original Off-Gas System 0 5 = 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

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! . D0ilBIT A t

j 3. SPECIFICATION 3 2.D, AIR FJECTOR OFF-GAS SYSTEM '

, REASON FOR OIANGE (continued)

he proposed changes would permit the air ejector monitor trip settin6 9 to be mised whenever the recombiners are in operation to a setting i based on the equivalent of the maximum pemitted stack release mte with a decay period of 120 minutes. 120 minutes cormsponds to the minimum expected Off-Gas System holdup time with the recombiners in opemtion.

With the recombiners inopemble and bypassed (Mode 3), the air ejector monitor trip setting; vould revert to the currently specified value.

Shifting from Modes 1 or 2 to Mode 3 requins a plant shutdown at which time the necessary trip setting changes vould be made.

The current Specification, to be placed in effect when the modifications to the Off-Gas System am completed and fully opemtional, is overly

} restrictive and cau lead to an unnecescary power reduction. Because of the higher ti.an avemF' rate of fuel clad defects at Monticello, the proportion of short-l'./ed radionuclides in the off-gas stream is nigher than normal. This shift in the distribution of the mixture of nuclides in the off-gas stream has only a slight effect on the stack release rate. We effect of the large fraction of short-lived

nuclides at the air ejector monitors, however, is to greatly increase the measured mdiation levels.

k. SPECIFICATION 3.8. A.3, MAXIMUM PERMI'PfED I-131 RELFASE RATt AND BASES t

PROPOSED QlANGE i

a. Change this Specification to read:

3. We maximum release mte of radioiodine 131 (I-131)shallnotexceedarateQ,in microcuries/ sect Q1 ,

QRS <y 40 27 -

b. Change the second pamgraph of the 3/h.8. A Bases on page 177A to state the correct location of the critical pathway dagry farm as 3700 meters in the. NNE sector stackX/Q=25x10- sec/m3and ground level X/Q = 4 3x10-7 sec/m ). Refer to the attached Exhibit B page 177A.

REASON FOR CHANGE Specification 3.8. A 3, to be placed it.to effect when the modifications to l the Off-Gas System are completed and fully operational, is now based on i the incorrect dairy fam location. Evaluation by NSP and the NRC Regulatory Staff has shown that the farm located 3700 meters from I-the site in the NNE sector constitutes the critical pathway.-

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E:0IIBIT A

.h. l

h. SPECIFICATION 3.8. A. 3, MAXDM4 PEM4I7 FED I-131 RELEASE RATE AND BASES REASON FOR CHAEiE (continued)

The proposed equstion is based on Regula.10 tory Guide 1.h2 and atmospheric dispersion factors calculated by the NRC Regulatory Staff.

5. SPECIFICATION 3/h.8.E, AUGMEIITED OFF-GAS SYSTD4 AND BASES PROPOSED CHANGE

a. Change Specification 3 8.E.2 to m adt

2. Except as specified in Specification 3 8.E.3 belov, at least one hydrogen monitor downstream of each operating recombiner shall be operable during power operation.

b. Change Specification 3.8.E.3 to readt

3. If the above specified downstream hydrogen monitors are not operable, offgas flov to the compressed storage subsystem shan be teminated.

c. Change Specification 4.8.E.2 to readt '

2. Condenser air inleakage shall be evaluated weekly and used in conjunction with the latest steam jet air ejector off-gas isotopic analysis and Figure h.8.1 to detemine that the limit of Specification 3 8.E.4 viu not be exceeded,

d. Delete Specification h.8.E.3

e. Add new Figure h.8.1, "Offgas Storage Tank Gross Activity Limits,"

as included on page 176A of Exhibit B.

f. Revise the 3/h.8.E Bases to reflect changes (a) through (e) above. Refer to pages 179A and 179B of Exhibit B.

REASON FOR CHANGE Change (a) deletes the requirement for one operable hydrogen monitor upstream of each operating recombiner. Planned modifications to the recombiner inlet flow control loops vill remove the upstream monitors.

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D0i! BIT A 5

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5. SPECIFICATION 3/h.8.E, AUGMENTED OFF GAS SYSTIN AND BASES REASON FOR CHANGE (continued)

Following modification, the inlet flow control loop vill be based on volumetric flow rate in lieu of hydrogen mass flow rate. Hydmgen concentration vill be assumed as the maximum design value at all times. This modification eliminates the requirement for inlet hydrogen measurement, inventory processing, and mass flow computation.

The Off-gas System has been operating since startup in a mode which-vill be made pemanent by the planned modification. Current 3y the inventory processors are programmed to assume a continuous hydrogen concentration equal to the design maximum in accordance with Specification 3.8.E.3 'Ibe effect of this mode of operation is to reduce the system flow capability significantly below the original.

design value, but not to the point of affecting normal plant operation.

Change (b) deletes the provision for allowing recombiner operation without an operable inlet hydrogen monitor. This provision vill no longer be needed when the upstream monitors are removed. Change (b) is also revised to require only temination of flow to the compressed storage subsytem when the required dovnetream hydmgen monitor is not operable. ' Ibis would pemit recombiner operation without requiring a return to the original Off-Gas System in the event that no hydrogen monitor is opemble. Other instrumentation vould be used to verify satisfactory operation of the recombiners.

All Off-Gas System components upstream of the compressed storage subsystem are designed to withstand a hydrogen detonation. The recombiner outlet hydrogen monitors serve to protect the compressed stomge subsystem components from a detonation that could result fmm excessive concentrations of hydrogen. If the downstream

, hydrogen monitors are inoperable, it is necessary to stop offgas

( flow to the compressed storage subsystem.

l Changes (c), (d), and (e) revise the method of comp 1,ying

! vith the maximum tank contents limit or 22,000 Curies dose i equivalent Xe-133 During startup testing of the augmented l Off-Gas System, it was found that the compressed storage tank radiation monitors do not perform their intended function.

L 'Ibe use of individual tank radiation monitors to measure gross l

radioactivity is.not feasible for the following reasons:

a. Each individual monitor is exposed to " shine" from adjacent storage tanks.

b. Each monitor becomes saturated due to the buildup of radioactive particulates (primarily Rb-88 and Cs-138 with high energy gamma radiation) and does not respond to changes in the noble-gas inventory of the tank.

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190!ZBIT A

5. SPECIFICATION 3/h.8.E, AUGMERfED OFF-GAS SYSTDI AND BASES REASON FOR CHANGE (continued)

Experience has also shown that the installed tank sampling system cannot be used to draw a representative sample of the contents of a tank while it is being filled. This is currently required by the Technical Specifications in the event a tank monitor is inoperable.

The location of the sample connections (in some cases on the tank fill lines) and stratification of gas have been founa to be responsible.

Calculations have been performed to determine the mlationship between air ejector off-gas activity and composition and consenser air inleakage. The results of these calculations are presented in Exhibit B Figure 4.8.1. A swanary of the technique used in performing these calculations is included as Appendix A to Exhibit A. .

It is pmposed that compliance with the 22,000 Curie dose equivalent Xe-133 tank contents limit be demonstrated by monitoring total system air inleakage and the average air ejector noble gas release mte. Using Figure h.8.1 and the results of the most recent detemination of noble gas isotopic composition at the air ejector, the contents of a storage tank can be verified to be below the activity limit.

Only under conditions of a high " equilibrium" type off-gas isotopic distribution, low condenser air inleakage, and high off-gas release mte is thers the potential for exceeding the 22,000 Ci limit.

As demonstrated in Appendix A to this Exhibit, Figure 4.8.1 is conservative.

6. TABLE OF CONTENTS AND LIST OF FIGURES PROPOSED CHANGE Replace pages y and v1 with the attached Exhibit B pages.

REASON FOR CHANGE This change revices the Table of Contents and List of Figures to include the new Specification 3/4.11 and the new Figures 3 6-1 through 3.6 4 and 4.8.1.

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, i APPENDIX A 20 DuiIBZT A LICENSE AMENDMEfff REQUEST DATED JULY 1,1975 NORTrERN STAT 13 PCMER COMPANY NUCLEAR SUPPORT SERVICES DEPARTMENT TECHNICAL REPORT OFFGAS STORAGE TANK C0fffEffPS FOR VARIOUS OFFGAS MIXTURES AND AIR INLEAYAGES AT THE M0fffICELID NUCLEAR GENERATING PLAlff

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,. Date: March 13, 1975 l .

l Reviced: June 5, 1975

1.0 Purpose Change 12 to the Monticello Technical Specifications issued by the USAEC Directorate of Licensing on November 15, 1973 placed a limit on the maioactivity that could be contained in an offgas storage tank of 22,000 curies dose equivalent Xe-133. me basis for this limit was accidental release of the contents of a tank by an operator ermr 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of filling. he release was assumed to occur over an eight hour period due to the flow restricting nozzle in the discharge line. Under accident 100-meter meteorology conditions, a whole body dose of 20 mrem would result to an individual at the site boundary. This dose corresponds to the Unusual Event annual gamma dose level.

To provide assurance that the 22,000 curie limit vould not be exceeded, it was believed that the tank radiation monitors could be calibrated in terms of gross gaseous activity. During startup testing, however, it was found that the presence of Ru-88 prevented the monitors from responding to gross gaseous activity changes. -

It has been su6gested that the tank radiation monitors may not be required. It should be possible to determine the maximum possible gross activity in a stomge tank by measuring total air inleakage, steam jet air ejector off-gas activity, and steam jet air ejector offgas isotopic composition. 2 determine the relationship between these parameters and tank activity in terms of dose equivalent Xe-133, the Monticello offgas system was modeled and a computer pmgram developed to perform the necessary calculations.

2.0 Besults_

Results of calculations of offgas storage tank content in terms of Xe-133 dose equivalent curies as a function of offgas isotopic distribution and total air inleakage are summarized in Tables 1 and 2 Table SJAE Monitor Trip Discharge Valve Setting Holdup Interlock Delay 1 30 min (current TS) 12 hr 2 2 hr (proposed TS) 12 hr Table 1 presents the results of the analysis for the current value of the steam jet air ejector monitor trip setting. Offgas is assumed to be released at the maximum permitted trip setting of I i 1 0.18, where kot is based on a decay time of 30 minutes. gamma'kot" The 22,000 curie limit is exceeded only at 3 C1H vith a 90%

equilibruim type distribution and at inleakages of less than 6 CFM with a 100% equilibrium distribution.

Table 2 presents the results of the analysis using the pmposed I me.ximum air ejector monitor trip setting which is based on a l decay time of 120 minutes. Again, the 22,000 curie limit is exceeded only at extreme combinations of lov inleakage and high equilibrium I

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, relations and actual IMR releases has been shown to be; good.

4 f -

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! All other delay times upstream of the compressed stora6e subsystem are l- 9.ssumed to be negligable.except the h2-inch pipe holdup.

l.

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oy:

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i

! Fifteen isotopec -vere considered in the model;(Table 2). htal release

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4 tot (t=30 min) Egamma' = 1.

) 0.18 l

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Ctot ( t ' ) = C(1, t' )

i=1,15 ,.

Cm33(t';= C(i,t')(Egammai/Egamma 13) i=1,15 -

1 Tank pressure as a function of filling time is:

p(psig) = (inleakage sofm)(t min)(14.69 psi)

(tank volume)

= L(sefm)t (min) 1h.69 neglecting temperature 1250 changes and the time to fill a tank becomes:

t f(min) = (285)(1250)

L(ecfm) 14.69 XE-133 dose equivalent activity in the tank after a delay of tintk when the tank discharge valve can be opened becomes therefore:

I

-(t d+tintk)Al -1t i y}

CxE133 " "Ii'0) # (1 ~

EQUIV g, 41 Egamma13.

-k.O Computer Program A Fortran program was written to perform the calculations outline in i

l Section 3 0. . Storage tank contents was calculated as a function of fill time and tank pressure for the following parameter variations:

a 1

Offgss distribution Recoil Fraction, FR 0 to 1.0, 0.1 steps Diffusion Fraction, Fp O to 1.0, 0.1 steps.

Equilibrium Fraction, FE O to 1.0, 0.1 steps Total air inleakage 3 to 30 cfm,1 cfm steps For each combination of offgas distribution and air inleakage, the program calculated the following quantities:

Release rate (uci/sec) Qtot (t) for t up to 600 hours0.00694 days 0.167 hours 9.920635e-4 weeks 2.283e-4 months

, Average gamma disintegration energy vs. t

! Average beta disintegration energy vs. t Q(1,t) for all 15 isotopes vs. t -

i C(1,t') for all 15 isotopes vs. t'

Ctot(t' )

CXE133(t')

EQUIV t'

f 4

j NormalizedQ(1,t)/yt1 % vs. %g Characteristic plot Program output for the 100% equilibrium distribution and 3 CFM total air inleakage is attached.

l

~

A series of additional calculations were also performed to determine the maximum possible tank activity for offgas release rates at the 4

' steam jet air ejector less than the maximum permitted trip setting.

Calculations were performed for air ejector monitor release rates

! of 10% to 100% of the maximum trip setting in 10% increments to establish the maximum permitted air ejector monitor release rate for any given air inleakage or offgas composition which will not exceed the 22,000 curie limit on tank activity. These results ny_ be used to determine an operating limit curve.

i l

l.

i l

l l !

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

Q( 1, t=ttot) , ,

A STACK 2.

XX PARTICUIATE j[

FILTERS 4 W I 4

4( Tm 5 )

j

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l~ J. CONDENSER .

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G O_- (X)MPRESSORS 6

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N CHARCOAL '

FILTER ;

( ---p- ) Q(1,t=td)

'y

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i' i

t FIGURE 1. OFFGAS SYSTD4 MODEL 1

1 1

r - = , , ,

TABLE 2. FISSION GASES CONSIDERED IN ANALYSIS NO. (i) ISOTOPE FISSION YIELD (%) DECAY CONSTANT (sec"1) E GAhMA (MEV/ die) 1 KR90 5 00 2.10E-2 2.10 2 XE139 5.40 1.69E-2 0.450 3 KRB9 h.59 3.61E-3 2.22 h XE137 6.00 2.96E-3 0.19h 5 XE138 5 90 8.1hE h 1.18 6 XE135M 1.80 7.222 4 0.k32 7 KBS7 2 53 1.f2E 4 0 793 8 KR83M 0 520 1.03E h 0.00248 9 KRB8 3 56 6.90E-5 1 95 10 KRB5M 1 30 4.38E-5 0.159 11 Xn35- 6.30 2.10E-5 0.247 la XE133M 0.160 3.h8E-6 0,0420 13 XE133 6.69 1 52E-6 0.0454 1h XE131M 0.0220 < 6.68E-7 0.0201 15 KB85 O.271 2.OhE-9 0.00220 Ref: ORNL-4723 & NEDO-10237 l TABLE 3 MODELING OF FIFSION OAS RELEASE RATE RELEASE bECHANISM MODEL RECOIL Q(i,t=0) = F3 Kgigi DIFFUSION Q(i,t=0) = F 9Kg141 !

EQUILIBRIUM Q( 1,t=o ) = FE Kg1 Q(1,t ) Release rate of fission gas i at i

time t F3,F3,FE Fraction of total offgas release attributed to recoil, diffusion, and equilibrium mechanisms l

K33 Kp,KE YF'er dependent constants 71 Fission fractional yield of fission gas 1 11 Decay constant of fission gas i Ref: GEI-92823A

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LICENSE AMENDMENT REQUEST DATED JULY 1,1975 ~
)
4 EXHIBIT B l-l 'Ihis exhibit consists of the following pages revised to incorporate the proposed Technical Specification changes:
I v
vi -
48
~~48A 115 116 i~~
116A (new page) 122 122A (new page 122B (Lev page i
122C (new page
, 130 131 131A (new pa6e) 169 173 173A
~~176A (new page) 177A 179A 179B -~~
~~189B (new page)~~
~~189c (new page 189D (new page 189E (new page~~
^
3 Standby Diesel Generators 182 ,
h. Station Battery Systems 183 39 Bases 185 h.9 Bases 186 3 10 and 4.10 Refueling 187 A. Refueling Interlocks 187
^
B. Core Monitoring 188 C. Fuel Storage Pool Water Level 188 D. Movement of Fuel 188 E. Bttended Core and Control Bod Drive Maintenance 188A 3 10 and 4.10 Bases . 189 3.n and 4.n Sealed Source contamination 189B 3.n and 4.n Bases 1895 50 DESIGN FEAWRES 190 6.O ADMINISTRATIVE COM'ROIS 192 6.1 Organization 192 6.2 Review and Audit 195 6.3 Actions to be taken in the Event of an Abnormal Occurrence 201 6.4 Action to be taken if a: Safety Limit is Exceeded 201 .
a
'65 Plant Operating Procedures- 202 6.6 Plant Operating Records 209
~~6.7 Plant Reporting Fequirements an V~~
' WE
LIST OF FIGURES Figure No. Page No.
2.1-1 Fuel Cladding Integrity Safety Limit 10 2.3 1 APRM Flow Referenced Scram and Rod Block Trip Settings 2 3 2. Relationship Between Peak Heat Flux and Power for Peaking Factor of 3 08 12 4.1.1 'M' Factor - Graphical Aid in the Selection of an Adequate Interval 46 Between Tests h.2.1 System Unavailability Th 3.4.1 Sodium Pentaborate Solution Volume - Concentration Bequirements 92 3.h.2 Sodium Pentaborate Solution Temperature Requirements 93
, 3.6.1 Change in Charpy V Transition Temperature versus Neutron Exposure 122 3.6.2 Minimum Temperature versus Pressure for Pressure Tests 122A 3.6.3 Minimum Temperature versus Pressure for Mechanical Heatup or 122B Cooldown Following Nuclear Shutdown '
3.6.h Minimum Temperature versus Pressum for Core _ Operation 122C h.6.1 Deleted h.6.2 Chloride Stress Corrosion Test Results @ 500 F 123
'! 4.8.1 Off-gas Storage Tank Gross Activity Limits 176A
. 6.1.1 ' NSP Corporate Organizational Relationship to On-site Operating Organization 193 6.1.2 Functional Organization for On-site Operating Group 194 vi REV
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS B. Emergency Core Cooling Subsystems Actuation When irradiated fuel is in the reactor vessel and the reactor water temperature is above ,
212 F, the limiting conditions for operation for the instrumentation which initiates the emergency core cooling sybsystems are given in Table 3.2.2.
C. Control Rod Block Actuation The limiting conditions of operation for the instrumentation that initiates control rod block are given in Table 3.2.3.
D. Air Ejector Off-Gas System
~~1. Except as specified in 3.2.D.2 and 3.2.D.3, both steam jet air ejector off-gas radiation monitors shall be operable during reactor power operation.~~
The trip settings for the air ejector monitors, except as specified in 3.2.D.4, shall be set to close the recombiner train inlet valve (s) within 30 minutes at a radiation level not to exceed the equivalent of the maximum permitted stack release rate after a decay time of 120 minutes.
32/h.2 L8 mv
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS
2. From and after the date that one of the two steam jet air ejector off-gas radiation monitors is made or found to be inoperable, continued reactor power operation is permissible ,
~~provided the inoperable radiation monitor -d instrument channel is tripped.~~
3. Upon loss of both steam jet air ejector off-gas radiation monitors, an orderly shut-down shall be initiated and the reactor shall be in cold shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
4. If operation is necessary with the Off-Gas Holdup System recombiners bypassed, the trip settings for the air ejector monitors shall be reset to close the stack off-gas isolation valve within 15 minutes at a radiation level not to exceed the equivalent of the maximum permitted stack release rate after a decay time of 30 minutes. x 3.2./4.2 48A REV
.j 3 0 LDGTING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIRDiENTS
[ 3.6 PRIMARY SYSTEN -BOUNDARY h.6 PRIMARY SYSTEM BOUNDARY
' Applicability: Applicability: ,
Applies to the operating status of the reactor Applies to the periodic eva=ination and testing "
l coolant system. requirements. for the reactor coolant system.
~~j. Objective: Objective:~~
To assure.the integrity and safe operation of the To detemine the condition of the reactor coolant reactor coolant system. system and the operation of the safety devices -t related to it.
1.
Specification: Specification:
i A. Reactor Coolant Heatup and Cooldown A. . Reactor Coolant Heatup and Cooldown 1-
~~1. . The average mte of reactor coolant During heatups and-cooldowns the following.~~
. temperature change during normal heatup temperatures shall be recorded..at least every or cooldown shall not exceed LOOO F/hr.
15 minutes until 3 consecutive mmngs at l
vhen averaged over a.one-hour period. each location are within 5 0F. E i
j a. Reactor vessel shell adjacent to shell flange.:
~~2. The pump in an idle recirculation loop b. Reactor vessel bottom' drain.~~
.'shall.not-be started unless the temper-ature of the coolant within the idle re- ., .c. Recirculation loops A and B.
, ciro11ation loop is vithin SOO F of the reactor coolant temperature. d. Reactor vessel bottom head.
I 36/4.6 4 115
. , . . ~~ _ _ _ .- - - _ ____
30 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMEITIS Reactor Vessel Temperature and Pressure B. Reactor Vessel Temperature und Pressure B.
1. During 'in-service hydrostatic or leak 1. During in-service hydrostatic or leak testing, the reactor vessel shell testing when the vessel pressum is temperatures specified in 4.6.B.1 shall above 312 psig, the following temper-be at or above the higher of the tem- atures shall be reconled at least every peratures shown on the two curves of 15 minutes.
Figure 3.6.2 where the dashed curve, "RPV Beltline Region," is increased a. Reactor vessel shell adjacent by the expected shift in RTygyp from to shell flange.
Figure 3.o.l.
~~b. Peactor vessel bottom head.~~
2. During heatup by non-nuclear means (except with the reactor vessel 2. Test specimens representing the vented), cooldown following nuclear reactor vessel, base veld, and veld shutdown, or lov level physics tests heat affected zone metal shall be the reactor vessel shell and fluid installed in the reactor vessel temperatures specified in 3.6.A shall adjacent to the vessel vall at the be at or above the higher of the core midplane level. The material temperatures of Figure 3.6 3 where the sample program shall conform to I dashed curve, "RPV Beltline Region," AS7M E 185-66. Samples shall be ;
~~is increased by the expected shift in withdrawn at one fourth and three RTygyp from Figure 3.6.1. fourths service life.~~
3. During all operation with a critical 3. Neutron flux vires shall be installed reactor, other than for low level in the reactor vessel adjacent to the physics tests or at times when the reactor vessel vall at the core mid-reactor vessel is vented, the reactor plane level. The wires shall be removed vessel shell and fluid temperatures and tested during the first mfueling specified in 3.6.A shall be at or outage to experimentally verify the above the higher of the temperatures calculated value of neutmn fluence at of Figure 3.6.4 where the dashed curve, one fourth of the beltline shell thickness "RPV Beltline Region," is increased that is used to determine the NI7IT by the expected shift in RT gyp from -
shift from Figure 3.6.1.
i Figure 3.6.1.,
116 36/h.6 ggy
30 LD4ITING CONDITIONS FOR OPERATION h.0 SURVEILIANCE REQUIRD4EITPS l+. The reac+,or vess21 head bolting studs 4. When the reactor vessel head studs are shall not be under tension unless the under tension and the reactor is in the tempe mture of the vessel head flange Cold Shutdown Condition, the reactor and the head ~are F(00F. vessel shell finnge temperature shall be pennanently recorded.
C. Coolant Chemistry C. Coolant Chemistry
1. The steady state radioiodine concentration 1. (a) A sample of reactor coolant shall be in the reactor coolant shall not exceed 5 taken at least every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months and microcuries of I-131 dose equivalent per grsm of water.
116A 36/h.6 REI
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1 Bases 3.6 and 4.6: A. Reactor Coolant Heatup and Cooldown The vessel has been analyzed for stresses caused by themal and pressure transients. Heating and cooling transients throughout plant life at unifom rates of 1000F per hour vere considered in the temperature range of 100 to Sh6oF and were shown to be within the requirements for stress intensity and fatigue limits of Section III of the ASME Boiler and Pressure Vessel Code. During reactor operation, the temperature of the coolant in an idle recirculation loop is expected to remain at reactor coolant temperature unless it is valved out of service. Requiring the coolant temperature in an idle loop to be within 500F of the reactor coolant temperature before the pump is started assures that the change in coolant temperature at the reactor vessel nozzles and bottom head region are within the conditions anal,yzed for the reactor vessel themal and pressure transients. B. Reactor Vessel Temperature and Pressure 1 Operating limits on the reactor vessel pressure and temperature during ncmal heatup and cooldown and during inservice hydrostatic testing were established using Appealix G of the Summer 1972
~~Addenda to Section III of the ASME Boiler and Pressure Vessel Code,1971 Edition, as a guide.~~
These_ operating limits assure that a large postulated surface flaw, having a depth of one-quarter of the material thickness, can be safely accommodated in regions of the vessel shell remote from discontinuities. For the purpose of setting these operating limits the reference temperature, RTmyf, of the vessel material was estimated from impact test data taken in acconlance with requirements of the Code to which this vessel was designed and manufactured (1965 Edition including Summer 1966 Addenda). Where the dropweight I&f temperature was known, the reference temperature used was the NDT temperature. Where the dropweight I&f temperature was not known, the reference temperature used was the temperature at which 30 ftib of energy was expected to occur on the basis of reported Charpy V notch test data. For areas of the vessel shell remote from the core beltline region, the highest I&I'I permitted by the vessel purchase specification for any vessel pressure boundary material is + hoof and this value is used for the RTmyf in lieu of certified test results.
'Ibe fracture toughness of all ferritic steels gradunlly and uniformly decreases with exposure to fast neutrons above a threshold value, and it is prudent and conservative 'to account for this in the operation of the reactor pressure vessel. Two types of infomation are needed in this nnalysis: a) A relationship between the change in fractum toughness of the reactor pressure vessel steel and the neutron fluence (integrated neutron flux),'and b) A measure of the neutron fluence at the point of interest in the reactor pressure vessel vall.
3.6A.6 BASES 130 REV
Bases 3.6 and 4.6 - continued: A relationship between neutron fluence and change in Charpy V notch test 30 ftib t m nsition temperature has been developed for SA302B/SA533 steel based on at least 35 experimental data points as shown in Figure 3.6.1. In turn this change in transition temperature can be related to a change in the temperature ordinate shown in Figure G 2110-1 in Appendix G of Section III of the ASME Boiler and Pressure Vessel Code. I The neutron fluence at any point in the pressure vessel vall can be computed from core physics data. The neutron fluence can also be measured experimentally on the inside diameter of the vessel vall. At present, valid experimental measurements can be made only over time periods of less than 5 years because of the limitations of the dosimeter materials. This causes no problem because of the exact relationship between themil power produced and the number of neutrons produced from a given core geometry. A single experimental measurement in a time period of one year can be used to predict the fluence for the life of the plant in thermal energy output if no great changes in core geometry are made. She vessel pressurization temperatures at any time period can be detemined from the thermal energy output of the plant and its relation to the neutron fluence and fmm Figure 3.6.1 used in conjunction with Figure 3.6.2 (pressure tests), Figure 3 6 3 (mechanical heatup or cooldown following nuclear shutdown), or Figure 3 6.4 (operation with a critical core). During the first fuel cycle, only calculated neutron fluence values can be used. At the first refueling, neutron dosimeter vires which are installed adjacent to the vessel vall are removed to verify the calculated neutron fluence. -- Fisin 3.6.1 viH be conservative for the Monticello reactor vessel. Reactor vessel material samples are provided, however, to verify the relationship expressed by Figure 3.6.1. Three sets of mechanical test specimens representing the base metal, veld metal, and weld heat affected zone (HAZ) metal have been placed in the vessel and can be mmoved and tested as required. These samples vill receive neutron exposure more rapidly than the vessel vall and therefore vill lead the vessel in integrated neutron flux exposure. An analysis and report vill be submitted to the Commission on all such surveillance specimens mmoved' from the reactor vessel in accordance with 10CFB50, Appendix H. 3hese reports shall include the information specified in AS3M E-185-66, " Recommended Practices for Surveillance Tests on Structural Materials in Nuclear Reactors," and information obtained on the level of integrated fast neutron irradiation received by the specimens and actual vessel material. 3.6/h.6 BASES 131 REV
A e-Bases 3.6 and h.6 - continued
'Ihe requirements for cold bolt-up of the reactor vessel closure are based on' the I&f temperature plus 600 F which is derived from the requirements of the ASME Boiler and Pressure Vessel Code to which the vessel was built. The IUf temperature of the closure flanges, adjacent head and shell naterial, and stud material is a maximum of 10 F. The minimum tempersture for bolt-up is therefore 100 + 600 = 7dOF. The neutron radiation fluence at the closure flanges is well below l@7 n/cm2 (E>l MEV) and therefore radiation effects vill.be minor and vill not influence this temperature.
t
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M J ' e i 3.6/4.6 BASES 131A ' wt 4 q ?. 1 4 6 v-- w
~
~~3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS~~
1. The maximum release rates of gross radio- 1. Radioactive gases released from the off-activity shall not exceed a rate Q, in gas stack and reactor building vent shall curies /sec: be continuously monitored. Station records of of f-gas stack release rates of gross q
Ey 4 [ Ev Ep }< gaseous radioactivity shall be maintaineo 0.18 \0.028 0.019/ - on an hourly basis to assure that the
2. Specified rates are not being exceeded, The release rates of gross radioactivity and to yield information concerning shall not exceed 16 percent of the limit general integrity of the fuel cladding.
~~in Specification 3. 8. A.1 avera ged over Records of isotopic analysis shall be any calendar quarter. maintained. The off-gas stack and~~
3. The maximum release rate of radioiodine reactor building vent monitoring system shall be functionally tested monthly 131 (1-131) shall not exceed a rate Q, in and calibrated quarterly with an appro-microcuries/sec: priate standard radiation source. Each monitor, as described, shall have a
~~+ < 1 sensor check at least daily.~~
4. The release rate of I-131 shall not exceed 2. A steam jet air ejector off gas sample 4 percent of the limit in Specification shall be taken and an isotopic analysis 3.8. A.3 averaged over any calendar quarter. for at least six fission product gases; Xe-138, Xe-135, Xe-133, Kr-88, Kr-85m,

5. The maximum release rates of radioactive Kr-87 shall be made at least weekly and particulates with half-lives greater than 8 following each refueling or other days shall not exceed a rate Q, in micro- occurrence which could alter significantly curies /sec: the mixture of radionuclides.
Q1 QRS 9.5x109 % + 2x108 HFCa 1 I whern FCh is the composite maximum per-missible concentration in air in uCi/ml determined using Appendix B Table 11, Column 1 and Notes of 10 CFR 20. 3.8/4.8 169 REV
3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS
3. Two independent samples of each tank shall 3. The performance and results of independent be taken and analyzed for gross beta-gamma samples and valve checks shall be logged.
~~activity and the valve line-up checked prior to discharge of liquid effluents.~~
~~4. If the limits of 3.8.C cannot be met, radio-active liquid effluents shall not be released.~~
D. Radioactive Liquid Storage D. Ra d ioa c tive Liquid Storage The maximum gross radioactivity in liquid storage 1. A sample shall be taken, analyzed, and in the Waste Sample, Floor Drain Sample, Waste recorded within 72 hours of each addition Surge, and Condensate Storage Tanks shall be to a liquid waste storage tank to which less than 30 curies except for tritium and Specification 3.8.D. applies. dissolved noble gases. If this condition cannot be met, the liquids in these tanks 2. If the sample analysis indicates that the shall be recycled to tanks within the radwaste total radioactivity in the liquid waste facility until the condition is met. storage tanks of Specification 3.8.D exceeds 30 curies, except for tritium and dissolved noble gases, the liquids in these tanks shall be recycled to reduce the radioactivity to less than 30 curies within 24 hours of this sampling. E. Augmented Of f-Gas System E. Augmented Of f-Gas System
~~1. If the hydrogen concentration in the off- 1.~~
gas downst ream of the recombiners reaches The hydrogen monitors shall be functionally tested monthly and calibrated quarterly four percent, the recombiner of f-gas flow with an appropriate gas mixture source. shall be stopped automatically by closing ' Each monitor shall have a sensor check the valves upstrean of the recombiners. at least daily.
2. Except as specified in Specification 2. Condenser air inleakage shall be 3.8.E.3 below, at least one hydrogen evaluated weekly and used in conjunction monitor downstream of each operating with the latest steam jet air ejector recombiner shall be operable during power operation.
off-gas isotopic analysis and Figure h.8.1 to determine that the limit of l Specification 3.8.E.h will not be exceeded. 3.8/6.8 173 REV
t 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3 If the above specified downstream hydrogen monitors are not operable, offgas flow to the compressed storage subsystem shall be temirated.
~~--]~~
~~4. The maximum gross radioactivity contained in one gas decay tank after 12 hours hold-~~
' up that can be discharged directly to the environs shall be less than 22.000 curies of Xe-133 dose equivalent. If these conditions cannot be met, the stored radioactive gas shall be recycled within 24 hours to other gas decay tanks until the condition is met.
5. During normal plant operation, radioactive gaseous vaste shall have a minimum holdup of 12 hours except for low radioactivity gaseous waste resulting from purge and fill operations associated with refueling and reactor startup. Holdup . times for radio-i active gaseous waste in the gas decay tanks shall be maximized consistent with plant operation.
F. Environmental Monitoring Program The environmental monitoring program given in Table 4.8.1 shall be conducted. 173A 3.8/4.8 REV L _ _ _ A_ . _ _ _ _
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#"^ !!E in e
3 8/h.8 176A RW
~
Bares Continued: Detailed meteorological calculations for several locations off site have been made by the AEC staf f and the most critical 22.5* sector was determined to be at 600 m to the south-southeast at the site boundary. The i annual average dif fusion parameter value for the of f gas stack release was determined to be 1.5 x 10-7 sec/m3
and for the reactor building vent release to be 7.2 x 10-6 sec/m3 i The method utilized by the staff to determine annual thyroid dose of 1500 mrem to a child for I-131 releases f rom the off gas stack and the reactor building vent is given in Regulatory Guide 1.42. Based on this method, the maximum I-131 concentration in milk f rom an existing cow would occur in the miE sector at a distance of 3700 m which has an annual average dif fusion parameter value of 2 5x10'O sec/m3 for the off-gas stack and h,3x10-7sec/m3 for the reactor building vent. Taking into account the five month grazir.g season a release rate of 1-131 from the off gas stack of 40 uCi/see or from the reactor butiding vent of 2.7 uC1/see co,ldu result.in an-annual thyroid dose of 1500 mrem to a child drinking this ellk.
In order to limit 1-131 releases in the gaseous effluents to as low as practical, quarterly average release rates have been established which would require investigative actions at 2 percent of the maximum release rate and plant actions at 4 percent of the maximum release rate. These release rates are significantly below 10 CFR Part 20 limits and are factors of 2 and 4, respectively, above the as low as practical objective of 1 percent of 10 CFR Part 20 limits. The AEC staff performed an analysis similar to that used to determine the maximum release rate of I-131 for the radioactive particulates with half-lives greater than 8 days. A reduction factor of 700 on the MPCa to allow for possible ecological chain effects similar to those associated with the cow-milk-child thyroid for radiciodine was used. The annual average diffusion parameters at 600 m in the south-southeast sector given previously were used for both the off gas stack and reactor building vent releases. Based on , these calculations, a continuous release rate of radioactive particulates with half-lives greater than 8 days in the amount of 9.5 x 109 MPCa uCi/sec from the off gas stack or 2 x 108 MFCa uCi/sec from the reactor building vent would not result in annual organ doses in excess of the limits specified in 10 CFR Part 20. < l In order to limit radioactive particulate releases in gaseous effluents to as low as practical, quarterly average release rates have been established which would require investigative actions at 2 percent of the maximum release rate and plant actions at 8 percent of the maximum release rate. These release rates are significantly below 10 CFR Part 20 limits and are factors of 2 and 8, respectively, above the as low as " practical objectives of 1 percent of 10 CFR Part 20 limits. 3.8/h.8 BASES 177A REV w r w'm 7
i i I t . 3 i i j Bases continued: . Each batch to be released will conform to 10 CFR Part 20 release limits on an instantaneous basis, i.e., I annual averaging will not be used as permitted by 10 CFR Part 20. See Section 9.2.3 of the FSAR. The radio- I activity. level in the discharge canal for a given release of waste will be the highest when the discharge
;. canal flow is lowest. This occurs during " closed cycle" cooling tower operation at which time the cooling tower blowdown of approximately 36 cubic feet per second is the major flow in the discharge canal. The
~~j. rate of psamping the radweste effluent into the discharge canal is variable and can, therefore, be controlled >~~
~~to maintain the concentration within the specified limit. This type of operatirn will be employed only~~
~~. when the river flow is very low and *will result in further dilution between distnarge canal effluent and~~
, the river. i . D. Radioactive Liquid Storage t The waste sample, floor drain sample, waste surge, and condensate storage tanks are not contained in a Class I ctructure. The === % = gross radioactivity in liquid storage in the specified tank = has been limited a on the basis of.an accidental spill from all stated tanka due to a seismic event great enough to damsge i them. Assuming a low recorded river flow of 1000 ft3 /sec, a day period over which the radioactive liquid wastes are diluted in the river, and consumption of the water by individuals at standard man consumption i rate (3000 al/ day), the single intake by an individual would not exceed one-third the yearly intake allowable by 10 CFR Part 20 for midentified radioisotopes (1 x 10-7 uCi/ml). The factor of 3 was applied to 10 CFR i Pcrt 20 limits as recommended for situations in which population groups could be exposed. ; The sampling frequency has been established so that if the maximum amount of gross radioactivity is exceeded, cetion can be taken to reduce the radioactivity to a level below the specified limit. E. Augmented Off-Gas System I l The hydrogen monitors are used to detect possible hydrogen buildups which could result in a possible hydrogen caplosion. Isolation of the off-gas flow would prevent the hydrogen explosion and possible damage to the 4
~~cugmented off-gas system. '~~
i ! j Experience has shown that a daily check with monthly testing and quarterly calibration assures proper operation l of the hydrogen monitors. The maxistan gross radioactivity in one gas decay tank has been limited on the basis that accidental release i cf its contents to the environs by operator error af ter 12 hours decay should not result in exceeding the I dose equivalent to the maximum quarterly release rate specified in Specification 3.8.A.2. Staff analysis of an elevated release under accident meteorology for a minime release period of 8 hours indicated a release of 22,000 curies of Xe-133 or the dose equivalent would result in a whole body dose of 20 mRen at
~~.the nearest site boundary.~~
3 8/4.8 BASES 179A l REV ; ) T
l l Bases Continued: Calculations and composition have andbeen perfomed condenser to detemine the relationship between steam jet air ejector off-gas acitivity air inleavage. These calculations were used to detem*_ne the curves presented in Figure 4.8.1. Se results of the weekly measurement of condenser air inleakage and the averege daily air ejector off-gas release rate are used in conjunction with the most mcent off-gas isotopic analysis to detemine if the maximum permitted Xe-133 dose equivalent tank radioactivity contents my be exceeded. Daily analysis is adequate , to detemine that if the mvin= amount of gross activity in a decay tank may be exceeded, action can be taken to reduce the radioactivity to a level below the specified limit, s F. Environmental Monitoring Program It is is recognized only possible by that a precise direct determination of environmental dose from a certain emission from the stack measurement. program conducted at and around the site.Such information vill be provided by the environmental monitoring I measureable in the environment, If the stack emission ever reaches a level such that it is limit long before the effect in the environmentsuch measurements vill provide a basis for adjusting the proposed stack is of any concern for permissible dose. In this regard, - it is important to realize that averaging emission rate over a period of one calendar year as permitted
by 10 CFR Part 20 represents a very large safety margin between conditions at any one instant (any minute, i hour, or day) and the long-term dose of interest.
~~i 1~~
3.8/h.8 bases 1798 t REY i i
i
i 1 l 4 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 ' SURVEILIANCE REQUIRH(ENTS l 3.11 SEALED SOURCE CONTAMINATION h.11 SEALED SOURCE CONTM4I14ATION Apulicability: Auplicability: Applies to each sealed source containing Applies to the periodic testing of more than 0.1 microcurie of plutonium or other special nuclear material (including alpha e-aled sources containic6 more than O.1 microcurie of plutonium or other radiation) and to each sealed source contain- special nuclear material (including ing more than the exempt quantities of alpha radiation) ani to each sealed bypn> duct materials listed in 107R30.71. source containing mre than the exe=pt quantities of byproduct materials listed in 10 7R30 71. Objective: Objective: To assure that leaka6e from sealed sources containing byproduct and special To verify the leak ti htness 6 of sealed nuclear radioactive materials does not radioactive sources. i exceed allowable limits. t O, Specification: Specification: A. Contamination A. Contamiret%
1. Each sealed source shall be free of 1. Tes* :_ for leakage and/or contamination removable contamination in excess of .aall be performed by the licensee or 0.005 microcuries Per 100'% smear test.
by other persons specifically authorized .' by the Occ=rission or an agreement State, as follows: 311/h.11 189B FE.T
9 30 IlMITING CONDITIOE FOR OPERATION h.O SUR7nW1 CE REQUIRECTIS
2. Each sealed source with re::nvable a . Each sealed source, except startup contamination in excess of the 11mit, ~
sources subject to core- flux, con-in 311.A.1 shall be 12:nediately with- taining Isdioactive material, other drawn from use and: than Hydztgen 3, with a half-life greater than 30 days and in any fom
~~a. Either decontaminated and repaired, other than p shn11 be tested for or .~~
~~leakage and/or contamtnation at~~
~~b. Disposed of in accordance with intervals not to exceed six months.~~
~~the regulations of the Ctmmission b . " die periodie leak test required does not apply to sealed sources that are stored and not being used. The sources~~
. exe=pted from this test shall be tested for leakage prior to any use . or transfer to another user unless . they have been leak tested within six nanths prior to the da'e_ of use or transfer. In the absence of a certificate from a transferor in-dicating that a test has been made within 1,1x months prior to the transfer, sealed sources shall not L
~~be put into use until tested for l leakage. l c . Startup sources shan be leak tested~~
~~' prior to sad following any repair or maintenance and before being subjected to core flux.~~
~~189C 311/h.11 Er/~~
~~- ~ - - - - - - - - - - - - - - -~~
~.
~~30 LntrrIm CONDITIONS FOR OPERATION h.O SURVEII. INK 2 REQUIREME!TIU~~
~~2. He leakage test shall be capable of detecting the presence of 0.005 microcuries of radioactive material per loofs s= ear test of the sa=ple.~~
~~B. Records~~
1. A complete inventory of maioactive materials in possession shall be unintained current at all times.

2. H e following records shall be n tained for two years:

a. Test results in microcuries, for tests performed pursuant to h.ll.A.

b. Eecord of annual physical inventory verifying accountability of sources on mcord.
~~3.uA.n 189D au~~
~~~. .~~
s s Bases 3 11 and h.11: The prograra, facilities, personnel, and procedures for safe stornge, handling, ani use of sealed sources containing radioactive m terials is described in Supplement No. 2 to the Application for Conversion of DPR-22 to Full Tem, submitted by Northern States Power Company on l August 16, 1974 The surveillance program described in these specifications is a part of the program to detect and control contamination of areas in the plant by such radioactive materials. h il . quantities of byproduct materials are exe=pt from licensing by 10 7330.18 ani therefore are exempt from leaks 6e tests in these specifications. Inhalation or injestion of such smn11 quantities of byproduct mterials from a sealed source vould result in less than one mrimm pemissible body burden for total body irradiation. Sources containing less than 0.1 micro-curie of plutonium are exempt from leakage tests by 10TETO.39(e) arai therefore . uch quantities of special nuclear sterials (inc1mHng alpha emitters) are exempt from leakage sests in thece specifications. The acceptance criteria of less than 0.005 microcurie on the test sample is also based on 107RTO.39(c). i 4 b 189E 3 11/h.11 BASES EE'i
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ML20125C513

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