ML20127K407

Amend 12 to License DPR-22 (Change 20) Changing TS Re Hydraulic Snubbers

ML20127K407
Person / Time
Site:	Monticello
Issue date:	09/15/1975
From:	Ziemann D Office of Nuclear Reactor Regulation
To:
Shared Package
ML20127K386	List: ML20127K382 ML20127K407 ML20127K414 ML20127K426
References
NUDOCS 9211200396
Download: ML20127K407 (13)

Text

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" U NIT E D STATES NUCLEAR REGULATORY COMMISSION W AS HIN GTO N D, C. 205 5 $

NORTHERN STATES PONER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING STATION AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 12 License No. DPR-22

1. The Nuclear Regulatory Commission (the Commission) has found that :

A. The application for amendment by Northern States Power Company (the .ticensee) dated August 15, 1975, complies with the stand'irds and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CIR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in corapliance with the Commission's regulations; and D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of

, the public.

2. Accordingly, the license is amended by a change to the Technical Specifications as indicated in he attachment to this license amendment and Paragraph 3.B ot ..icense No. DPR-22 is hereby amende ' to read as follows:

"B. Technical Specifications

• The Technical Specifications contained in Appendix A as revised, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications, as revised by issued changes thereto -hrough Change No. 20."

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9211200396 750915 PDR ADOCK 05000263 ,

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3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COSNISSION Orid"* L

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Dennis..

Dennis L. Zicmann, Chief Operating Reactors Branch #2 Division of Reactor Licensing 4

Attachment:

Change No. 20 to the Technical Specifications Date of Issuance:

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! i j ATTACINENT TO LICENSE AMENDMENT NO.12.

CllANGE NO. 20 TO Ti1E TECIINICAL SPECIFICATIONS PROVISIONAL OPERATING LICENSE NO. DPR-22 DOCKET NO. S0-263 Replace the existing pagas of the Technical Specifications listed below with attached revised nages bearing the same nwabers, except as other-wise noted. Changen o. these pages are shown by marginal lines.

Pages lii vii 121~

121A - Add 121B Tabic 3.6.1, page 1 of 4 - Add 121C Tabic 3.6.1, page 2 of 4 - Add 121D Tabic 3.6.1, page 3 of 4 - Add 121E Table 3.6.1, page 4 of 4 - Add 138 l' 138A - Add i

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D. 11PCI Syste

n 103 .

E. , Autoestic Pressure Relief Syste:s 104 F. RCIC System 106 i

G. Minimu:s Core and Contair=ent Cooling Systen Availability 107 H. Deleted ,

I. ~Recirculaeion Syatem 108A i

! - J. . Average Planar UlGR 108A i K. Local UIGR 1088 3.5 Bases 109  !

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4.5 Esaes ,

114 ,

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j 3.6 and 4.6 Primary System Boundary 115

A. Thermal Limitations. 115 B.. Pressurization Temperature 116 C. Coolant' Chemistry -

116 1

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{ D. Coolant Irskage 118 i i

! E. Safety / Relief valves 119 1 l I .,

F. Structural Integrity 120 - .!

l . G.: Jet Puc:ps 120 Ilydraulic Snubbers 20 3 11. 121  ;

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j 3.6 and 4.6 Bases 130 t

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4 1.lST OF TABLES I

• i 3.1.1 Fcactor Protection System (Scram) Inst ruacnt Requirements 30 ,

l 4.1.1 Scram Instrument. Functional Tests - Minimum Functional Test Frequencias 34 for Safety. Instrumentation and Control Circuits 1

4.1.2 Scram Instrument Calibration - Minimuu Calibration Frequencies for 36 Reactor Protection Instrument Channels i

3.2.1 Instrumentation that Initiates Primary Containment Isolation Functions 50

'3.2.2 Instrumentation that Initiates Emergency Core Cooling 53 Systems 3.2.3 Instrumentation that Initiates Rod Blwk 57 ,

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3. 2. l ' Instrumentation that Initiates Reactor fluilding Ventilation 60 ,

t Isolition and Standby Gas Treatment System Initiation 3.2.5' ' Trip Functions and Deviations 69 .;

-4.2.1 Minimum Test.and Calibration. Frequency For Core Cooling, Rod Block 61 i and Isolation Instrumentation ,

.3.6.1 Safety Related Ilydraulic Snubbers 121B 20

4.6.1 -In-Service Inspection Requirements for Monticello 124

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3.7.1 Primary Containnient lsolation 153 4 4.8.1 Sample Collection and Analysis Monticello Nuclear Plant.- Environmental 174 j

Monitoring Program '

-6.5.1 Protection Factors for Respirators 206 1

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3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILIANCE REQUIREMENTS

3. The diffuser to lower plenum differential pressure reading on an individual jet pump is 107. or more, less than the mean of all jet pump differentia pressures.

11 . Hydraulic Snubbers

11. 11ydraulic Snubbers -

1. During all modes of operation, except Cold Shut-down and Refueling Shutdown, all hydraulic The folleuing surveillance requirements apply to all hydraulic snubbers listed in Table 3.6.1:

snubbers listed in Table 3.6.1 shall be operable E

except as noted in 3.6.11.2 through 3.6.11.4 i

below.

1. All hydraulic snubbers whose seal material has been demonstrated by operating experience, lab testing

2. From and after. the time that a hydraulic stubber or analysis to be compatible with the operating is detemined to be inoperable, continued reactor environment shall . be visually inspected to veri fy operation is permissible only during the succeed- their operability in accordance with the folicwing ing 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless the snubber is sooner made schedule:

operable.

Number of Snubbers Next Required Found Inoperable Inspection 5 3. If the requirements of 3.6.11.1 and 3.6.11.2 cannot surint, Inspection be met, an orderly shutdown shall be initiated Interval or During Inspectica and the reactor shall be in a cold shutdown Interval condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

0 18 months + 257.

4. If a hydraulic snubber listed in Table 3.6.1 is determined to be inoperable while the reactor 1

12 months I 257.

2 6 months I 257.

, is in the shutdown or refueling mode, the snubber 3,4 124 days + 257.

shall be made operable prior to reactor startup. 5,6,7 62 days [+257.

>8 31 days + 257.

The required inspection interval shall not be lengthened more than one step at a' time.

j 3.6/4.6 121

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3.0 LIMITING' CONDITIONS FOR OPERATION 4.0 SURVEILIECE REQUIREMENTS

5. Snubbers may be added to safety related snubbers may be categorized in two groups, systems without prior License Amendment " accessible" or " inaccessible" based on their to Table 3.6.1 provided that safety accessibility for inspection during reactor  !

cvaluations, documentation and reporting operation. These two groups may be inspected 20 _

are provided in accordance with IOCfU independently according to the above schedule. '

50.59 and that a revision to Table ,'1 -

is included with a subsequent Licen' 2. All hydraulic snubbers whose seal materials have Amendment request. not been demonstrated to be compatible with the operating environment shall be visually inspected for operability every 31 days. ,

3. The initial inspection shall be performed within I

6 menths frina the date of issuance of these specifications. For the purpose of entering the schedule in Specification 4.6.11.1 it shall be assumed that the facility had been on a 12-month

, inspection interval.

4. Once each refueling cycle, a representative saaple of 10 snubbers or approximately 10% of the snubbers, whichever is less, shall be functionaat) tested for operability including verification of proper piston movement, lock up and biced. For each unit and subsequent unit found inoperabic, 1

i an additional 10% or ten snubbers shall be so tested until no more failures are found or all units have been tested.

5. Once each refueling cycle at least tuo repre-sentative snubbers frem a relatively severe l environment shall be completely disassembled and  ;

1 examinedfordamageandabnormalsealdegradation.I 3.6/4.6 121A

i TABLE 3.6.1 (Page 1 of 4 )

Safety Related Hydraulic Snubbers .

Accessible or Inaccessible Snubber No. Location - System - Elevation (A or I)

SS-1 Drywell - Main Steam 953' Az 279 I SS-2, Drywell - Main Steam 953' Az 810 1 SS-3 Drywell - Main Steam 950' Az 2120 I ,

SS-4 Drywell - Main Stean 950' Az 1480 I l SS-7 Drywell - Main Steam 953' Az 240') I SS-8 Drywell - Main Steam 953' Az 1200 I SS-11 Drywell - Feedwater 952' Az 3020 1 SS-12 Drywell - Feedwater 952' Az 380 I i

. 20 SS-13 Drywell - Feedwater 952 Az 2580 I i SS-14 Drywell - Feedwater 952' Az 960 I SS-17A Drywell .- lillR 964' Az 720 1 SS-17B Drywell - RilR 964' Az 720 I [

SS-18A ' Drywell - Ri!R 964' Az 28S0 ,

I j SS-18B Drywell - R!lR -

964' Az 2880 ' ~

i SS-19 Drywell - RIIR 964' A: 3410 1 1

SS-20 'Drywell - RIIR 964' Az 19 I I

SS-1AR Drywell - Recirculation 922' Az 315o I

.SS-1BR Dryuell--. Recirculation 922' Az 135o I  !

SS-2AR Drywell - Recirculation 927' Az 3020 I i

SS-2BR Drywell - Recirculation 927' A: 1220 1 >

, SS-3AR Drywell - Recirculation 927' Az 3280 I SS-3BR Drywell - Eccirculation 927' Az 1480 I SS-4AR(a) Drywell - Recirculation 934' Az 3020 I l , SS-4AR(b) Drywell - Recirculation 934' Az 3230 I SS-41:R (a) Drywell - Recirculation 934' Az 1200 1 SS-4BR(b) Drywell - Recirculation 934' Az 1490 1 SS-SAR Drywell - Recirculation'. 941' Az 315 1 SS-5BR Drywell - Recirculation 941' Az 135 I ,

a SS-6AR Drywell - Recirculation 953' Az 261 1 SS-6BR Drywell - Recirculation 953' Az 990 I 3.6/4.6 121B i

TABLE 3.6.1 (Page 2 of 4 )

Safety Related Ilydraulic Snubbers Accessible or Inaccessible Snubber No. I.ocation - System - Elevation (A or I)

SS-7AR Drywell - Recirculation 953' Az 323 I SS-7BR Drywell - Recirculation 953' Az 320 1 SS-8AR Drywell - Recircitlation 927' Az 2700 I SS-8BR Drywell - Recirculation 927' Az 900 I PSI-ll2 Drywell - Main Steam 953' Az 710 I PS L-Il3 Drywell - Main Steam 950' Az 1480 I PS2-Il2 Drywell - Main Steam 950' Az 1200 I PS3-II2 Drywell - Main Steam 950' Az 2400 I PS4-il3 Dryuell - Main Steam 950' Az 2120 1 20- RV24-il3 Drywe11 - Safety Re1ief 950' Az 1100 I RV24-il4 Dryuell - Sa fety Relief 935' Az 1000 I RV24-II4A Dryuell' - Sa fety Relie f 935' Az 100') I RV24-Il5 Drywell - Safety Relief 935' Az 1100 I RV24A-Il4A ~ Drywell - Safety Relief 957' Az 4 80 I RV24A-Il7 'Drywell - Safety Relief 953' Az 1150 I RV24A-!!8 Drywell - Safety Relief 939' Az 32 I RV25-ill Drywell '- Safety Relief 953' Az 1800 I RV25-Illa Drywell - Safety Relie f 953' Az 1800 I RV25-II2 Drywell - Safety Relief 948' Az 1900 I RV25-ll2A Drywell - Safety Relief 948' Az 1900 I RV25-ll3 Drywell -' Sa fety Relie f 934' Az 1800 I i RV25A-II2 Drywell - Safety Relief 945' Az 1200 I RV25A-112A Dryuell - Safety Relief 945' Az 120' I s RV25A-II7 Dryuell - Sa fe ty. Relie f 953' Az 1350 I RV26-ill Dryvell - Sa fety ReIle t 953' Az 200) I RV26-Illa Dryuell .- Sa fety Relie f 953' Az 200) 1 RV26-II2 Drywell .- Safety Relief 947' Az 20P I RV26-il2A Drywell - Safety Relie f 947' Az 20m I RV26A-Il2 - Drywell - Sa fety Relie f 940' Az 250) I RV26A-II2A Dryuell - Safety Relle t 935' Az 250) I 3.6/4.6 121 c u

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TABLE 3.6.1 (Page 3 of 4 )

Safety Related Hydraulic Snubbers 1 Accessibic or Inaccessli)e Snubber No. Location - System - Elevation (A or I)

RV27,-Ill Drywell . Safety Relief 950' Az 320 1 RV27-IIIA Drywell - Safety Relief 950' Az 230 0 1 RV27-Il5 Drywell - Safety Relier' 945' Az'270 I RV27-II6 Drywell - Safety Relief 945' Az 270 I

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' RV27A-Il2A- Drywell - Safety Relief 953' Az 290 I RV27A-Il3 Drywell - Safety Relief 953' Az 290 I RV27A-119 Dryvell - Safety Relief 938' Az 290 1 SS-21 Torus Floor Level - RilR South Wall A

SS-22 Torus Floor Level - R!la South Wall A SS-23 "B" RilR. Room - RilR Floor Level A SS-24 "A" EllR Room - lutR Floor Level A SS-25 RIIR Discharge - !!!In Southeast wall just , A 20 -

belou torus catwalk SS-26 "B" KllR Room - Core Spray Floor Level A

SS-27 "B" RIIR Room - Core Spray Floor Level A

SS-28A "A" RilR Room - Core Spray Floor Level A SS-28B "A" RIIR Room - Core Spray Floor. Level A SS-19 Overhead, by N2 Analyzer - RIIR 954' A 4

SS-30 -Overhead, by N2 Analyzer - RIIR 954' A

SS-31 Torus catwalk - RilR Discharge ---

A

SS-32A "A" RilR Room; Behind IIeat Exchanger- RIIR 916' A j SS-32B "A" R11R . Room; Behind IIcat Exchanger- RIIR 916' A 4

, SS-33 Above Torus on side sloping towards

- Drywell - RllR Discharge ---

A l SS-34 Above Torus on side sloping towards i Dryuel1 - IlllR Discharge ---

A SS-35 IIICI Room - IIPCI Pump Discharge On North Wall, 912'_ A SS-36A IIPCI Room - IIPCI Turbine D<haust Floor Level A SS-36B  !!PCI Room - IIPCI Turbine D:haust Floor' Level A 4

3.4/4.6 121 o

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t T/ .'E 3.6.1 (Page 4 of 4 )

Sa fe t s.  ; elated llydraulic Snubbers Accessible or 4

Inaccessible Snubber No. Location - System - Elevation (A or I)

I l.. SS-37 IIPCI Room - IIPCI Turbine i>haust West Wall, 905' A i SS-38A 'RCIC Room - RCIC Turbine lhhaust IJest Wall, 906' ,

A SS-38B RCIC Roem - RCIC Turbine '>.haust West Wall, 906' A 20 SS-40 Main Stean Chase - IIPCI S. naa supply ---

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SS-41 Above Torus Catwalk - Ct,re S tiray Discharge 927' A SS-42 Above Torus Ring IIeader I!PCI Steam Exhaust North West Wall, 906' ~A N

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Bases Continued 3.6 and 4.6: s A nozzle-riser system' failure could also generate the coincident failure of a jet pump body; however, the l converse is not true. The lack of any substantial stress in the jet pump body makes failure impossible without an initial nozzle-riser system failure.  !

II . Ilydraulic Snubbers 1,

Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earth-quake or severe transient, while allowing nortual thermal motion during startup and shutdown. The consequence '

, of an inoperabic snubber is an increase in the probability of structural damage to piping as a res2 7t of a seismic or.other event initiating dynamic loads. It is therefere required that all hydraulic snubbers required .!

to, protect the primary coolant system or any other safety system or component be operable during reactor operation.

20 Because the snubber protection.is required only during relati'vely low probability events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for repairs or replacements. In case a shutdown is required, the allevance of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to reach a cold ,

shutdown condition will permit an orderly shutdown consistent with standard operating procedures. Since plant startup should not commence with knowingly defective safety related equip = cut, Specificatien 3.6.11.4 prohibits startup with inoperable snubbers.

All safety related hydraulic snubbers are visually inspected for overall integrity and operability. The in-spection will include verification of proper orientation, adequate hydraulic fluid level and prpper attachment of snubber to piping and structures. -

t The inspection frequency is based upon maintaining a constant level of snubber protection. Thus the required i inspection interval varies inversely with the observed snubber failures. The number of inoperabic snubbers found during a' required inspection determines the time interval for the next required inspection. Inspections

performed-before that interval has elapsed may be used as a new reference point to determine the next in-

spection. Ilowe ver , the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule

' Experience at. operating facilities has shoun that the required surveillance program should assure an acceptable level of snubber .perfor, nance provided that the seal materials are compatible with the eperating environment.

Snubbers containing seal material which has not been demonstrated by operating experience, lab tests or analysis to be compatible with the operating environment should be inspected more frequently (every month) until material compatibility is confirmed or an appropriate changcout is completed. '

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II. Ilydraulic Snubbers (contd.)

Examination of defective snubbers at reactor facilities and matecial tests performed at several laboratories has'shown that millable gum polyurethanc deteriorates rapidly under the temperature and moisture conditions present in many snubber locations. Although molded polyurethane exhibits greater resistance to these conditions, it also may be unsuitable for application in the higher temperature environments. Data are not currently available to precisely define an upper terperature limit for the molded polyurethane. Lab tests and in-plant experience indicate that seal materials are available, primarily ethylene propylene compounds, which should give satisfactory performance under the most severe conditions expected in reactor installations.

To further increase the. assurance of snubber reliability, functional tests should be performed once each l refueling cycle. .These tests will include stroking.of the snubbers to verify proper piston movement, lock--

up and bleed. Ten percent or ten snubbers, whichever is less, represents an adequate sample for such tests.

] Observed failures on these samples should require testing of additional units. Snubbers in high radiation

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areas or those especially difficult to remove need not be selected for functional tests provided' operability

.was previously verified. To ccmplement the visual external inspections, disassembly and internal examination i

for component damage and abnomal seal degradation should be performed. The examination of two units, each a

refueling. cycle, selected from relatively severe environments shculd adequately serve this purpose. Any t

observed wear, breakdown or deterioration .uill provide a basis for additional inspections.

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3.6/4.6 BASES 138A I

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