Text

Amends 85 & 78 to Licenses DPR-19 & DPR-25,respectively, Revising Tech Specs Re Change in Installation or Use of Facility Components Located within Restricted Area,Per 10CFR20 & Change to Surveillance Requirements
	ML17346B341
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Site:	Dresden
Issue date:	02/27/1985
From:	Zwolinski J; Office of Nuclear Reactor Regulation
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)ty**4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 COMMONWEALTH EDISON COMPANY DOCKET NO. 50-237 DRESDEN NUCLEAR POWER STATION, UNIT NO.

2 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 85 License No.

DPR-19 1.

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

A.

The application for amendment by the Commonwealth Edison Company (the licensee) dated March 15, 1984 as supplemented by a letter dated September 21, 1984 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility wi 11 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 wil'1 be conducted in compliance with the Commission's regulations; 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; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 nf the Commission's regulations and all applicable requirements have been satisfied.

8503070403 850227 PDR ADOCN 05000237,'

PDR

0 Cl f j 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.8 of Provisional Operating License No.

DPR-19 is hereby amended to read as follows:

B.

Technical S ecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 85, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR TH NUCLEAR REGULATORY MNI ION

Attachment:

Changes to the Technical Specifications John

. Zwolinski, Chief Opera

'ng Reactors Branch PS Division of Licensing Date of Issuance:

February 27, 1985

ATTACHMENT TO LICENSE AMENDMENT NO.

85 PROVISIONAL OPERATING LICENSE OPR-19 COCKET NO. 50-237 Revise Appendix A Technical Specifications by removing the paqes identified below and inserting the attached pages.

The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE 1iii(1) vii(1) 3/4.6-12 3/4.6-13 3/4.6-15 through 3/4.6-24 B3/4.6-25 through B3/4.6-38 1NSERT 1iii(2) vii(2) 3/4.6-12 3/4.6-13 3/4.6-15 through 3/4.6-20 B3/4.6-21 through B3/4.6-34

( 1) Paces issued by Amendment 83 (RETS).

(2) Pages which supersede those issued by Amendment 83 (Amd. 83 does not become effective until March 15, 1985).

DRESDEN II DPR-3,9 Amendment No. g, p5~

8~

APPEND'IX h TO OPERATING LICENSE DPR-19 TECHNICAL SPECIFICATIONS AND BASES FOR DRESDEN NUCLEAR POWER STATION UNIT 2 GRUNDY COUNTY ILLINOIS COMMONWEALTH EDISON COMPANY DOCKET NO. 50-237 3918a 8401D

DRE N II Amendment No.

DPR-19 y6, fi,, 85 (Table of Contents Cont'd.)

~Pa e

3.5.C 3.5.D 3.5.E 3.5.F 3.5.G 3.5.H 3.5.I 3.5.J 3.5.K 3.5.L 3.6 3.6.A 3.6.B 3.AC 3.6.D 3.6.E 3.6.F 3.6.G 3.6.H 3.6.I 3.7 3,7.A 3.7,B 3.7.C 3.7.D 3.8 3.8.A 3.8.B 3.8.C 3.8.D 3.8.E 3.8.F 3 '

3.9.A 3.9.B stem Availability (3.5)

(3.6)

(3.7) ces (3.8)

HPCI Subsystem Automatic Pressure Relief Subsystems Isolation Condenser System Minimum Core and Containment Cooling Sy Deleted Maintenance of Filled Discharge Pipe Average Planar LHGR Local LHGR Minimum Critical Power Ratio Condensate Pump Room Flood Protection Limiting Conditions for Operation Bases Surveillance Requirement Bases (4')

Primary System Boundary Thermal Limitations Pressurization Temperature Coolant Chemistry Coolant Leakage Safety and Relief Valves Structural Integrity Jet Pumps Rec ircu1 at ion Pump Flow Mism at ch Shock Suppressors (Snubbers)

Limiting Conditions for Operation Bases Surveillance Requirement Bases (4.6)

Containment Systems Primary Containment Standby Gas Treatment System Secondary Containment Primary Containment Isolation Valves Limiting, Conditions for Operation Bases Surveillance Requirement Bases (4.7)

Radioactive Materials Airborne Effluents Mechanical Vacuum Pump Liquid Effluents Radioactive Waste Storage General Information Miscellaneous Radioactive Materials Sour Limiting Conditions for Operation Bases Surveillance Requirement Bases (4.8)

Auxiliary Electrical Systems Requirements Availability of Electric Power 3/4. 5 3/4. 5 3/4.5 3/4.5 3/4.5 3/4.5 3/4.5 3/4.5 3/4.5 B 3/4.5 B 3/4.5 3/4.6 3/4.6 3/4.6 3/4.6 3/4.6 3/4.6 3/F 6 3/4.6 3/4.6 3/4.6 B 3/4.6 B 3/4.6 3/4.7 3/4.7 3/4.7 3/4.7 3/4.7 B 3/4.7 B 3/4.7 3/4.8 3/4 '

3/4.8 3/4.8 3/4.8 3/4.8 3/4.8 B 3/4.8 B 3/4.8 3/4.9 3/4.9 3/4.9 6

8 9

-11

-13

-15

-25

-26

-31

-39 1

1 2

3 5

6 7

-10

-11

-12

-21

-34 1

1

-19

-25

-27

-33

-40 1

1 9 ll

-12

-14

-18

-22 1

- 1 2

3959a 3843A

DRESDEN II DPR-19 Ament No.

P2, List of Tables Table 3.1.1 Table 4.1.1 Table 4.1.2 Table 3.2.1 Table 3.2.2 Table 3.2.3 Table 3.2.4 Table 3.2.5 Table 4.2.1 Table 4.2.2 Table 4.2.3 Table 4.6.2 Table 3.7.1 Table 4.8.1 Table 4.8.2 Table 4.8.3 Table 4.8.4 Table 4.8.5 Table 4.8.6 Table Table Table Table Table Table 3.12-1 3.12-2 3

~ 1 2 3

3. 12-4 6.1.1 6.6.1 Table 4.11-1 Reactor Protection System (Scram)

Instrumentation Requirements Scram Instrumentation Functional Tests Scram Instrumentation Calibration Instrumentation that Initiates Primary Containment Isolation Functions Instrumentation that Initiates or Controls the Core and Containment Cooling System Instrumentation that Initiates Rod Block Radioactive Liquid Effluent Monitoring Instrumentation Radioactive Gaseous Effluent Monitoring Instrumentation Minimum Test and Calibration Frequency for Core and Containment Cooling Systems Instrumentation, Rod Blocks, and Isolations Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Neutron Flux and Sample Mithdrawal Primary Containment Isolation Radioactive Gaseous Waste Sampling and Analysis Program Maximum Permissible Concentration of Dissolved or Entrained Noble Gases Released From the Site to Unrestricted Areas in Liquid 'Haste Radioactive Liquid Haste Sampling and Analysis Program Radiological Environmental Monitoring Program Reporting Levels for Radioactivity Concentrations in Environmental Samples Practical Lower Limits of Detection (LLD) for Standard Radiological Environmental Monitoring Program Surveillance Requirements for High Energy Piping Outside Containment Fire Detection Instruments Sprinkler Systems C02 Systems Fire Hose Stations Minimum Shift Manning Chart Special Reports

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3/4. 1 5

3/4.1 8

3/4.1 -10 3/4.2 - 8 3/4 '

-10 3/4.2 -12 3/4.2 -14 3/4.2 -15 3/4.2 -17 3/4.2 -20 3/4.2 -22 B 3/4.6-26 3/4.7 -31 3/4.8-22 3/4.8-24 3/4.8-25 3/4.8-27 3/4.8-28 3/4.8-29 3/4.11-3 B 3/4.12-17 B 3/4.12-18 B 3/4.12-19 B 3/4.12-20

&, 21 6-5 6-26 vil 3959a 3843A

DRESDEN II DPR-19 dmendment Nn. P4 3.6 LIMITING CONDITION FOR OPERATION (Cont'd.)

4.6 SURVEILLANCE RE UIREMENT (Cont'd.)

Whenever both recirculation pumps are in operation, pump speeds shall be maintained within 1W of each other when power level is greater than 80% and within 15%

of each other when power level is less than 807.

Recirculation pumps speed shall be checked daily for mismatch.

2. If specification 3.6.H.1 cannot be met, one recirculation pump shall be tripped.

3.

The reactor shall not be operated with one recirculation loop out of service for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With the reactor operating, if one recirculation loop is out of service the plant shall be placed in a hot shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless the loop is sooner returned to service.

4, Whenever one pump is operable and the remaining pump is in the tripped position, the operable pump shall be at a speed less than 657. before starting, the inoperable pump.

I.

Snubbers (Shock Suppressors)

I.

Snubbers (Shock)

Suppressors)

The following surveillance requirements apply to safety related snubbers.

3/4.6-12 3688a

0 DRESDEN II 3g R-19 Amendment No.

3.6 LIMITING CONDITION POR OPERATION (Cont'd.)

4.6 SURVEILLANCE RE UIREMENT (Cont'd,)

1.

During all modes of operation except cold shutdown and refuel, all safety related snubbers shall be operable except as noted in Specifi-cation 3.6.I.2 through 3.6.I.4-.

l.

Uisual Inspection An independent visual inspection shall be performed on the safety related hydraulic and mechanical snubbers in accordance with the schedule below.

a.

All hydraulic snubbers whose seal material has been demonstrated by operating experience, lab testing or analysis to be compatible with the operating environment shall be visually inspected.

This inspection shall include, but not necessarily be limited to, inspection of the hydraulic fluid reservoir, fluid connections, and linkage connection to the piping and anchor to verify snubber operability.

b.

All mechanical snubbers shall be visually inspected.

This 3/4.6-13 3688a 3123A

DRESDEN II RPR-19 Amendment No.

3.6 LIMITING CONDITION FOR OPERATION (Cont'd.)

4.6 SURVEILLANCE RE UIREMENT (Cont'd.)

2.

From and after the time e snubber is determined to be inoperable, continued reactor operation is permissible only during the succeeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months unless the snubber is sooner made operable or replaced.

2.

Functional Testing a.

Once each refuel-ing cycle, a

representative sample of approxi-mately 107 of the hydraulic snubbers shall be functionally tested for operability, including:

(i)Activation (restraining ection) is achieved within the specified range of velocity or acceleration in both tension and compression.

(ii)

Snubber

bleed, or release

rate, where required, js within the specified range in compression or tension.

3/4.6-15 3688a

DRESDEN II QFR-19 Amendment No.

3.6 LIMITING CONDITION FOR OPERATION (Cont'd.)

4.6 SURVEILLANCE RE UIREMENT (Cont'd.)

For each unit and subsequent unit found inoperable, an additional 10%

of the hydraulic snubbers shall be tested until no more failures are found or all units have been tested.

b.

Once each refueling

cycle, a

representative sample of "approximately 10%,

of the mechanical snubbers shall be functionally tested for operability.

The test shall consist of two parts:

(i)

Verification. that the force that initiates free movement of the snubber in either tension or compression is less than the specified maximum breakaway friction force.

3/4.6-16 3688a 3123A

DRESDEN II DPR-19 Amendment No. g

~ 85 3.6 LIMITING CONDITION FOR OPERATION (Cont'd.)

4 '

SURVEILLANCE RE UIREHENT (Cont'd.)

(ii) Verify that the actxvatxon (restraining action) is achieved within the specified range of acceleration or

velocity, as applicable based on snubber design in both tension and compression.

For each unit and subsequent unit, found inoperable, an additional 10% of the mechanical snubbers

. shall be so tested until no more failures are found or all units have been tested.

c, In addition to the regular sample, snubbers which failed the previous functional test shall be retested during the next test period.

If a spare snubber has been installed in place of a failed

snubber, then both the failed snubber (if it is repaired and installed in another position) and the spare snubber shall be retested.

Test results of these snubbers may not be included for the resampling.

3/4.6-17 3688a 3123h

E. ~

'I c

DRESDEN II DPR-19 Amendment No. P

~ ~5 3.6 LIMITING CONDITION FOR OPERATION (Cont'd.)

4.6 SURVEILLANCE RE VIREMENT (Cont'd.)

3. If the requirements of 3.6.I.1 and 3.6.1.2 cannot be

met, an orderly shutdown shall be initiated and the reactor shall be in cold shutdown or refuel condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

3.

When a snubber is is deemed inoperable, a review of all pertinent facts shall be conducted to determine the snubber mode of failure and to decide if an engineering evaluation should be performed on the supported system or components.

If said evaluation is deemed necessary, it will determine whether or not the snubber mode of failure has imparted a

.significant effect or degradation on the supported component or

system,

4. If a snubber is determined to be inoperable while the reactor is in the cold shutdown or refuel mode, the snubber shall be made operable or replaced prior to reactor startup.

4. If any snubber selected for functional testing either fails to lock up or fails to move, i.e., frozen in

place, the cause will be evaluated and, if determined to be a

generic deficiency, all snubbers of the same design subject to the same defect shall be functionally tested.

3/4.6-18 3688a

DRESDEN II DPR-19 Amendment No, g

~ ~

3.6 LIMITIHG CONDITION FOR OPERATION (Cont'd.)

4.6 SURVEILLANCE RE UIREMENT (Cont'd.)

5.

Snubbers may be added or removed from safety related systems with-out prior license amendment.

5.

Snubber service life monitoring shall be followed by existing station record systems, including the central filing system, mainten-ance files, safety re-lated work packages, and snubber inspection records.

The above record retention methods shall be used to prevent the hydraulic snubbers from exceeding a service life of 10 years and the mechanical snubbers from exceeding a

service life of 40 years (lifetime of the plant).

3/4.6-19 3688a 3123A

ey DRESDEN II QPR-19 kaendmeet Ne. pd lan)rntrrrr T~altrre Rrrqviremrnts per Apprrnr)ix G ol 'Io CF R 50 900 Z

~.

Cr loo zoo

$00 S

S00 K

400 0

300 CURVE A IINSERVICE PRESSURE TESTS-SECTION XI) 1 CURVE B (HEATUP ~ COOLOOWN)

CURVE C (CfllTICALCORE OPERATION) 0 MINIMUMBOLTING TEMPFRATURE 100 F

MINIMUMOPERATING TEMPERATURE 149 I'-

RTNOT ~ 40 F

Kl PER SECTION G'21IOOF APPENO X G OF THE SUMMER )973 AOOENOA TO SECTION III OF THE ASME COOE TEMPERATURE

( F)

Fig. 3.6.1 MINIMUM TEMPERATURE REQUIREMENTS PER APPENDIX G OF 10 CFR 50 3/4.6-20 3688a

DRESDEN II DPR-19 Amendment No. g

~ 85 3.6 LIMITING CONDITION FOR OPERATION BASES A.

Thermal Limitations The reactor vessel design specification requires that the reactor vessel be designed for a maximum heatup and cooldown rate of the contained fluid (water) of 100'F per hour averaged over a period of one hour.

This rate has been chosen based on past experience with operating power plants.

The associated time periods for heatup and cooldown cycles when the 100 F per hour rate is limiting provides for efficient, but safe, plant operation.

The reactor vessel manufacturer has designed the vessel to the above temperature criterion.

In the course of completing the

design, the manufacturer performed detailed stress analysis.

This analysis includes more severe thermal conditions than those which would be encountered during normal heating and

cooling operations.

Specific analyses were made based on a heating and cooling rate of 100 Flhour applied continuously over a temperature range of 100'F to 550 F.

Because of the slow temperature-time response of the massive flanges relative to the adjacent head and shell sections, calculated temperatures obtained were 500 F (shell) and 360 F (flange)

(140 F differential).

Both axial and radial thermal stresses were considered to act concurrently with full primary loadings, Calculated stresses were within ASME Boiler and Pressure Vessel Code Section III stress intensity and fatigue limits even at the flange area where maximum stress occurs.

The flange metal temperature differential of 140 F occurred as a result of sluggish temperature response and the fact that the heating rate continued over a 450'F coolant temperature range.

The uncontrolled cooldown rate of 240'F was based on the maximum expected transient over the lifetime of the reactor vessel.

This maximum expected transient is the injection of cold water into the vessel by the high pressure coolant injection subsystem.

This transient was considered in the design of the pressure vessel and five such cycles were considered in the design.

Detailed stress analyses were conducted to assure that the injection of cold water into the vessel by the HPCI would not exceed ASME stress code limitations.

B.

Specification 3.6.A.4 increases margin of safety for thermal-hydraulic stability and startup of recirculation pump.

B 3/4.6-21 3688a

DRESDEN II QPR-19 Amendment Ne. )V2 3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

Pressurization Tem erature The reactor coolant system is a

primary barrier against the release of fission products to the environs.

In order to provide assurance that this barrier is maintained at a high degree of integrity, restrictions have been placed on the operating conditions to which it can be subjected.

These restrictions on inservice hydrostatic

testing, on heatup and cooldown, and on critical core operation shown in Figure 3.6,1, were established to be in conformance with hppendix G to 10 CFR 50.

In evaluating the adequacy of ferritic steels Sa302B it is necessary that the following be established:

a)

The reference nil-ductility temperature (RTNDT) for all vessel and adjoining materials, b) the relationship between RTNDT and integrated neutron flux (fluence, at energies greater than one Mev),

and c) the fluence at the location of a postulated flow.

The initial RTNDT of the main closure flange, the shell and head materials connecting to these flanges, and connecting welds is 10 F.

However, the vertical electroslag welds which terminate immediately below the vessel flange have an RTNDT of 40 F, (Reference hppendix F to the FSAR)

The closure flanges and connecting shell materials are not subject to any appreciable neutron radiation exposure, nor are the vertical electroslag seams.

The flange area is moderately stressed by tensioning the head bolts.

Therefore, as is indicated in curves (a) and (b) of Figure 3.6.1, the minimum temperature of the vessel shell immediately below the vessel flange is established as 100'F below a pressure of 400 psig.

(40'F

+

60'F, where 40'F is the RTNDT of the electroslag weld and 60 F is a conservatism required by the hSME Code).

hbove approximately 400 psig pressure, the stresses associated with pressurization are more limiting than the bolting stresses, a

fact that is reflected in the non-linear portion of curves (a) and (b).

Curve (c), which defines the temperature limitations for critical core operation, was established per Section IV Z.c. of hppendix G of 10CFR50.

Each of the curves, (a),

(b) and (c) define temperature limitations for unirradiated ferric steels.

Provision has been made for the modification of these curves to account for the change i,n RTNDT as a result of neutron embrittlement.

B 3I4.6-22 3688a

I I

0 DRESDEN II PPR-19 Amendment No.

3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.}

The withdraw'al schedule in Table 4.6.2 is based on the three capsule surveillance program as defined in Section 11.C.3.a of 10 CFR 50 Appendix H.

The accelerated capsule (Near Core Top Guide) is not required by Appendix H but will be tested to provide additional information on the vessel material.

This surveillance program conforms to ASTH E 185-73 "Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels" with one exception.

The base metal specimens of the vessel were made with their longitudinal axes parallel to the principal rolling direction of the vessel plate.

C, Coolant Chemistr A radioactivity concentration limit of 20 Micro-Ci/ml total iodine can be reached if the gaseous effluents are near the limit as set forth in Specification 3.8.C.1 or there is a failure or a prolonged shutdown of the cleanup demineralizer.

In the event of a steam line rupture, outside the drywell, the resultant radiological dose at the site boundary would be about 10 rem to the thyroid.

This dose was calculated on the basis of a total iodine activity limit of 20 Micro-Ci/ml, meteorology corresponding to Type F

conditions with a one meter, per second wind speed, and a valve closure time of five seconds.

If the valve closed in ten

seconds, then the resultant dose would increase to about 25 rem.

The reactor water sample will be used to assure that the limit of Specification 3.6.C is not exceeded.

The total radioactive iodine activity would not be expected to change rapidly over a

period of 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months .

In addition, the trend of the stack off-gas release

rate, which is continuously monitored, is a

good indicator of the trend of the iodine activity in the reactor coolant.

Since the concentration of radioactivity in the reactor coolant is not continuously measured, coolant sampling would be ineffective as a means to rapidly detect gross fuel element failures.

However, some capability to detect gross fuel element failures is inherent in the radiation monitors in the off-gas system and on the main steam lines.

Materials in the primary system are primarily 304 stainless steel and the Zircaloy fuel cladding.

The reactor water chemistry limits are established to prevent damage to these materials.

Limits are placed on chloride concentration and conductivity.

The most important limit is that placed on chloride concentration to prevent stress corrosion cracking of 3688a 3123A B 3/4.6-23

DRESDEN II DPR-19 Amendment No. g

~ 85 3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

the stainless steel.

The attached graph, Figure 4.6.2, illustrates the results of tests on stressed 304 stainless steel specimens.

Failures occurred at concentrations above the curve; no failures occurred at concentrations below the curve.

According to the data, allowable chloride concentrations could be set several orders of magnitude above the established limit, at the oxygen concentration (0.2-0.3 ppm) experienced during, power operation.

Zircaloy does not exhibit similar stress corrosion failures.

However, there are various conditions under which the dissolved oxygen content of the reactor coolant water could be higher than 0.2-0.3

ppm, such as refueling, reactor startup and hot standby.

During these periods with steaming rates less than 100,000 pounds per hour, a more restrictive limit of 0.1 ppm has been established to assure the chloride-oxygen combinations of Figure 4.6.2 are not exceeded.

At steaming rates of at least 100,000 pounds per hour, boiling occurs causing deaeration of the reactor water, thus maintaining oxygen concentration at low levels.

When conductivity is in its proper normal range, pH and chloride and other impurities affecting conductivity must also be within their normal range, When and if conductivity becomes

abnormal, then chloride measurements are made to determine whether or not they are also out of their normal operating values.

This would not necessarily be the case.

Conductivity could be high due to the presence of a neutral salt; e.g.,

Na2S04, which would not have an effect on pH or chloride.

In such a case, high conductivity alone is not a

cause for shutdown.

In some types of water-cooled reactors, conductivities are in fact high due to purposeful addition of additives.

In the case of BWR's, however, where no additives are used and where neutral pH is maintained, conductivity provides a very good measure of the quality of the reactor water.

Significant changes therein provide the operator with a warning mechanism so he can investigate and remedy the condition causing the change before limiting conditions, with respect to variables affecting the boundaries of the reactor

coolant, are exceeded.

Methods available to the operator for correcting the off-standard condition include, operation of the reactor clean-up

system, reducing the input of impurities and placing the reactor in the cold shutdown condition.

The major benefit of cold shutdown is to reduce the temperature dependent corrosion rates and provide time for the clean-up system to re-establish the purity of the reactor coolant.

B 3l4. 6-24 3688a 3123h

t

)

~

DRESDEN II DPR-19 Amendment.

No. g, 85 o<o

+

I

~I Ol 0%

~o TYPICAL CURVE,

<0 lo'S SO" NTECRATEO REUTROII EXPOSURE ) I WV~)

loll Figure 4,6.1 MINIMUM REACTOR PRESSURIZhTION TEMPERATURE B 3/4.6-25 3688a

DRESDEN II PPR-19 hmendment No. g~

~>

ThBLE 4.6.2 NEUTRON FLUX hND ShMPLES MITHDRhVAL SCHEDULE FOR DRESDEN UNIT 2 Withdrawal Year Part No.

Location Comments 1977 Near Core Top Guide 180 hccelerated Sample 1980 Mall 215'000 7

9 10 Mall 95'all

245'all 275 Standby Standby B 3/4.6-26 3688a

( I y l

DRESDEN II PPR-19 hmendment No. pl IOO lo FAILURE 5

l.o O.l Llatl EIIahlilhcd I<<CI al OOerlting O> Ccncea0<<ion = 0.0 -O.l @pa IIO FAILURE 0.1 I.O REFERENCE; CORROSIOR 0 IIEAR IIAROOOOK O. J. Oehwl, Ed.

IO CHLORIDE Od<<l Figure 4.6.2 CHLORIDE STRESS CORROSION TEST RESULTS hT 500 F

B 3/4.6-27 3688a

DRESDEN II QPR-19 Amendment No. g~

~>

3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

During start-up periods, which are in the category of less than 100,000 pounds per hour, conductivity may exceed 2

micro-mho/cm because of the initial evolution of gases and the initial addition of dissolved metals.

During, this period of time, when the conductivity exceeds 2 micro-mho (other than short term spikes),

samples will be taken to assure the chloride concentration is less than 0.1 ppm.

The conductivity of the reactor coolant is continuously monitored.

The samples of the coolant which are taken every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months will serve as a reference for calibration of these monitors and is considered adequate to assure accurate readings of the monitors. If conductivity is within its normal range, chlorides and other impurities will also be within their normal ranges, The reactor coolant samples will also be used to determine the chlorides.

Therefore, the sampling frequency is considered adequate to detect long-term changes in the chloride ion content.

Isotopic analyses required by Specification 4.6.C.3 may be performed by a gamma scan.

D.

Coolant Leaka e

Allowable leakage rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes and on the ability to makeup coolant system leakage in the event of loss of offsite a-c power.

The normally expected background leakage due to equipment design and the detection capability for determining coolant system leakage were also considered in establishing the limits.

The behavior of cracks in piping systems has been experimentally and analytically investigated as part of the USAEC sponsored Reactor Primary Coolant System Rupture Study (the Pipe Rupture Study).

Work utilizing the data obtained in this study indicates that leakage from a crack can be detected before the crack grows to a dangerous or critical size by mechanically or thermally induced cyclic loading, or stress corrosion cracking or some other mechanism characterized by gradual crack growth.

This evidence suggests that for leakage somewhat greater than the limit specified for unidentified leakage, the probability is small that imperfections or cracks associated with such leakage would grow rapidly.

However, the establishment of allowable unidentified leakage greater than that given in 3.6.D on the basis of the data presently available would be premature because of uncertainties associated with the data.

For leakage of the order of 5

gpm as specified in 3.6.D, the experimental and analytical data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation.

Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available B 3/4,6-28 3688a

DRESDEN II QPR-19 amendment No. P4 3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd,)

leakage dete'ction

schemes, and if the origin cannot be determined in a reasonably short time the plant should be shut down to allow further investigation and corrective action.

The additional leakage requirements will be in effect only while the reactor is operated with the recirculation flaws detected during the 1983 Refueling Outage, The additional leakage requirements will provide more conservative detection and corrective action should the current flaws propagate thru wall.

The capacity of the drywell sump is 100 gpm and the capacity of

~

the drywell equipment drain tank pumps is also 100 gpm, Removal of 50 gpm from either of these sumps can be accomplished with considerable margin.

The performance of reactor coolant leakage detection system will be evaluated during the first five years of station operation and the conclusions of this evaluation will be reported to the NRC.

It is estimated that the main steam line tunnel leakage detection system is capable of detecting the order of 3000 lb/hr.

The system performance will be evaluated during the first five years of plant operation and the conclusions of the evaluation will be reported to the NRC.

E.

Safet and Relief Valves - The frequency and testing, requirements for the safety and relief valves are specified in the Inservice Testing Program which is based on Section XI of the ASME Boiler and Pressure Vessel Code.

Adherence to these code requirements

'rovides adequate assurance as to the proper operational readiness of these valves.

The tolerance value is specified in Section III of the ASME Boiler and Pressure Vessel Code as plus or minus 1% of design pressure, An analysis has been performed which shows that with all safety valves set 1% higher than the reactor coolant pressure safety limit of 1375 psig is not exceeded.

The safety valves are required to be operable above the design pressure (90 psig) at which the core spray subsystems are not designed to deliver full flow.

F.

Structural inte rit - A pre-service inspection of the components in the primary coolant pressure boundary will be conducted after site erection to assure the system is free of gross defects and as a reference base for later inspections.

Prior to operation, the reactor primary system will be free of gross defects.

In addition, the facility has been designed such that gross defects should not occur throughout life.

B 3/4.6-29 3688a

DRESDEN II PPR-19 Amendment No.

3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

lnservice Inspections of ASME Code Class 1,

2 and 3 components will be performed in accordance with the applicable version of Section XI of the ASME Boiler and Pressure Uessel Code.

Relief from any of the above requirements must be provided in writing by the Commission.

The Inservice Inspection program and the written relief do not form a part of these Technical

'pecifications.

These studies show that it requires thousands of stress cycles at stresses beyond any expected to occur in a reactor system to propagate a crack.

The test frequency established is at inter-vals such that in compari,son to study results only a small num-ber of stress

cycles, at values below limits will occur.

On this basis, it is considered that the test frequencies are adequate.

The type of inspection planned for each component depends on lo-cation, accessibility, and type of expected defect.

Direct visual examination is proposed wherever possible since it is sensitive, fast and reliable.

Magnetic particle and liquid pene-trant inspections are planned where practical, and where added sensitivity is required'ltrasonic testing and radiography shall be used where defects can occur on concealed surfaces'fter five years of operation, a program for in-service inspec-tion of piping and components within the primary pressure boundary which are outside the downstream containment isolation valve shall be submitted to the NRC.

G.

~yet Pum s

Patlure oy a jet pump nozzle assembly hold down mechanism, nozzle assembly and/or riser increases the cross sectional flow area for blowdown following the postulated design basis double-ended recirculation line break.

Therefore, if a failure occurs, repairs must be made to assure the validity of the calculated consequences.

The following factors form the basis for the surveillance requirements:

A break in a jet pump decreases the flow resistance character-istic of the external piping loop causing the recirculation pump to operate at a higher flow condition when compared to previous operation.

The change in flow rate of the failed jet pump produces a change in the indicated flow rate of that pump relative to the other pumps in that loop.

Comparison of the data with a normal relationship or pattern provides the indication necessary to detect a failed jet pump.

B 3I4.6-30 3688a

g

~

DRESDEN II DPR-19 hmendment No. P2 3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

The jet pump flow deviation pattern derived from the diffuser to lower plenum differential pressure readings will be used to further evaluate jet pump operability in the event that the jet pumps fail the tests in Section 4.6.G.1 and 2.

Agreement of indicated core flow with established power-core flow relationships provides the most assurance that recirculation flow is not bypassing the core through inactive or broken jet pumps.

This bypass flow is reverse with respect to normal jet pump flow.

The indicated total core flow is a

summation of the flow indications for the twenty individual jet pumps.

The total core flow measuring instrumentation sums reverse jet pump flow as though it were forward flow.

Thus the indicated flow is higher than actual core flow by at least twice the normal flow through any backflowing pump.

Reactivity inventory is known to a high degree of confidence so that even if a jet pump failure occurred during, a shutdown

period, subsequent power ascension would promptly demonstrate abnormal control rod withdrawal for any power-flow operatng map point.

h nozzle-riser system failure could also generate the coincident failure of a jet pump body; however, the 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.

H.

Recirculation Pum Flow Mismatch The LPCI loop selection logic has been described in the Dresden Nuclear Power Station Units 2 and 3 FSAR, Amendments 7

and 8.

For some limited low probability accidents with the recirculation loop operating with large speed differences, it is possible for the logic to select the wrong loop for injection.

For these limited conditions, the core spray itself is adequate to prevent fuel temperatures from exceeding allowable limits.

However, to limit the probability even

further, a procedural limitation has been placed on the allowable variation in speed between the recirculation pumps.

The licensee's analyses indicate that above 80% power the loop select logic could not be expected to function at a speed differential of 157, Below 807. power, the loop select logic would not be expected to function at a speed differential of 20%,.

This specification provides a margin of 5V in pump speed differential before a problem could arise.

If the reactor is operating on one pump, the loop select logic trips that pump before making the loop selection.

B 3/4.6-31 3688a

DRESDEN II DPR-19 Amendment No. g

~ ">

3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

In addition; during the start-up of Dresden Unit 2, it was found that a flow mismatch between the two sets of jet pumps caused by a difference in recirculation loops could set up a

vibration until a mismatch in speed of 27%. occurred.

The 107 and 15% speed mismatch restrictions provide additional margin before a pump vibration problem will occur.

ECCS performance during reactor operation with one recirculation loop out of service has not been analyzed'herefore, sustained reactor operation under such conditions is not permitted.

I.

Snubbers (Shock Su ressors)

Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient while allowing normal thermal motion during startup and shutdown.

The consequence of an inoperable snubber is an increase in the probability of structural damage to piping as a result of a seismic or other event initiating dynamic loads.

It is therefore required that all snubbers required to protect the primary coolant system or any other safety system or component be operable during reactor operation.

Because the snubber protection is required only during low probability events, a period of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed for repairs or replacements.

In case a shutdown is required, the allowance of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to reach a cold shutdown condition will permit an orderly shutdown consistent with standard operating procedures.'ince plant startup should not coaxnence with knowingly defective safety related equipment, Specification 3.6.I.4 prohibits startup with inoperable

snubbers, Mhen a snubber is found inoperable, a review shall be performed to determine the snubber mode of failure.

Results of the review shall be used to determine if an engineering evaluation of the safety-related system or component is necessary.

The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the support component or system, hll safety related hydraulic snubbers are visually inspected for overall integrity and operability.

The inspection will include verification of proper orientation, adequate hydraulic fluid level and proper attachment of snubber to piping and structures.

B 3/4.6-32 3688a

DRESDEN II PPR-19 Amendment No.

3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

All safety related mechanical snubbers are visually inspected for overall integrity and operability.

The inspection will include verification of proper orientation and attachments to the piping and anchor for indication of damage or impaired operabili ty.

The inspection frequency is based upon maintaining a constant level of snubber protection.

Thus, the required inspection interval varies inversely with the observed snubber failures.

The number of inoperable 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 inspection.

However, 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.

To further increase the assurance of snubber reliability, functional tests will be performed once each refueling cycle.

A representative sample of 10% of the safety-related snubbers will be functionally tested.

Observed failures on these samples will require testing of additional units.

Hydraulic snubbers and mechanical snubbers may each be treated as different entities for the above surveillance programs.

Hydraulic snubber testing will include stroking of the snubbers to verify piston movement, lock-up, and bleed.

Functional testing of the mechanical snubbers will consist of verification that the force that initiates free movement of the snubber in either tension or compression is less than the maximum breakaway friction force and verification that the activation (restraining action) is achieved within the specified range of acceleration or velocity, as applicable based on snubber

design, in both tension and compression.

B 3/4,6-33 3688a 3123A

3,6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

Shen the cause of rejection of the snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible, and verified by inservice functional testing, that snubber may be exempted from being counted as inoperable.

Generically susceptible snubbers ere those which ere of a specific make or model and have the same design features directly related to rejection of the snubber by visual inspection or are similarly located or exposed to the same environmental conditions such as temperature, radiation, and vibration.

Monitoring of snubber service life shall consist of the existing station record systems, including the central filing

system, maintenance files, safety-related work packages, and snubber inspection records.

The record retention programs employed at the station shall allow station personnel to maintain snubber integrity.

The service life for hydraulic snubbers is 10 years.

The hydraulic snubbers existing locations do not impose undue safety implications on the piping and components because they are not exposed to excesses in environmental conditions.

The service life for mechanical snubbers is 40 years, lifetime of the plant.

The mechanical snubbers are installed in areas of harsh environmental conditions because of their dependability over hydraulic snubbers in these areas.

All snubber installations have been thoroughly engineered providing, the necessary safety requirements.

Evaluations of all snubber locations and environmental conditions justify the above conservative snubber service lives'.

4.6 sURvEILLANcE RE UIREMENT BAsEs None 3688e B 3/4.6-34

~P,g AEVI Wp O~

o

+%*4+

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 COMMONWEALTH EDISON COMPANY.

DOCKET NO. 50-249 DRESDEN NUCLEAR POWER STATION, UNIT NO.

3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 78 License No.

DPR-25 1.

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

A.

The application for amendment by the Commonwealth Edison Company (the licensee) dated March 15, 1984 as supplemented by a letter dated September 21, 1984 complies with standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR 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 compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the coImIIon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's requlations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by charges to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 3.B of Facility Operating License No.

DPR-25 is hereby amended to read as follows:

B.

Technical S ecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.

78, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR TH UCLEAR REGULATORY OMMI ION John

. Zwolinski, Chief Operat'ng Reactors Branch ¹5 Division of Licensing

Attachment:

Changes to the Technical Specifications Date nf Issuance:

February 27,

]9Q5

~

I

ATTACHMENT TO LICENSE AMFNDMENT NO.

78 FACILITY OPERATING LICENSE DPR-25 DOCKET NO. 50-249 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.

The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of chanae.

REMOVE vii(1) viii(1) 3/4.6-12 3/4.6-13 3/4.6-15 through 3/4.6-24 B3/4.6-25 thorugh B3/4.6-38 INSERT 1 11~

vii((2) viii(2) 3/4.6-12 3/4.6-13 3/4.6-15 through 3/4.6-20 B3/4.6-21 through B3/4.6-34

( 1)

Pages issued by Amendment 77 (RETS).

(2) Pages which supersede those issued bv Amendment 77 (Amd.

77 does Rot become effective until March 15, 1985).

DRESDEN III Amendment No.

j77, DPR-25 P,

7S (Table of Contents Cont'd.)

~Pa e

3 '.C 3.5.D 3.5.E 3.5.F 3.5.G 3.5.H 3.5.I 3.5.J 3.5.K 3.5.L 3,6 3.6.A 3.6.B 3.6.C 3.6,D 3.6.E 3.6.F 3,6.G 3.6.H 3,6.I 3.7 3.7.A 3.7.B 3.7.C 3.7,D 3.8 3.8.A 3.8.B 3.8.C 3.8.D 3.8.E 3.8.F 3.9 3.9.A 3.9.B stem Availability (3.5)

(3.6)

(3.7) rces (3.8)

HPCI Subsystem Automatic Pressure Relief Subsystems Isolation Condenser System Minimum Core and Containment Cooling Sy Deleted

'aintenance of Filled Discharge Pipe Average Planar LHGR Local LHGR Minimum Critical Power Ratio Condensate Pump Room Flood Protection Limiting Conditions for Operation Bases Surveillance Requirement Bases (4.5)

Primary System Boundary Thermal Limitations Pressurization Temperature Coolant Chemistry Coolant Leakage Safety and Relief Valves Structural Integrity Jet Pumps Recirculation Pump Flow Mismatch Shock Suppressors (Snubbers)

Limiting Conditions for Operation Bases Surveillance Requirement Bases (4.6)

Containment Systems Primary Containment Standby Gas Treatment System Secondary Containment Primary Containment Isolation Valves Limiting Conditions for Operation Bases Surveillance Requirement Bases (4.7)

Radioactive Materials Airborne Effluents Mechanical Vacuum Pump Liquid Effluents Radioactive Waste Storage General Information Miscellaneous Radioactive Materials Sou Limiting, Conditions for Operation Bases Surveillance Requirement Bases (4,8)

Auxiliary Electrical Systems Requirements Availability of Electric Power 3/4. 5 3/4.5 3/4.5 3/4.5 3/4. 5 3/4. 5 3/4. 5 3/4. 5 3/4. 5 3/4. 5 3/4.5 3/4. 6 3/4. 6 3/4. 6 3/4. 6 3/4. 6 3/4. 6 3/4. 6 3/4. 6 3/4. 6 3/4. 6 3/4, 6 3/4. 6 3/4, 7 3/4. 7 3/4. 7 3/4, 7 3/4

~ 7 3/4.7 3/4.7 3/4.8 3/4.8 3/4.8 3/4.8 3/4.8 3/4.8 3/4.8 3/4.8 3/4.8 3/4.9 3/4 '

3/4.9 6

8 9-ll

-13

-15

-22

-23

-29

-37 5

6 7 ll

-12

-21

-34 1

1

-19

-25

-27

-33

-40

- 1 1

9

-10

-11

-12

-14

-18

-22

- 1

- 1 2

3958a 0009A

List of Tables DRE~ III DPR-25 AmeniKent No. g, g~

78

~Pa e

Table 3.1.1 Table 4.1

~ 1 Table 4.1.2 Table 3.2.1 Table 3.2.2 Table 3.2.3 Table 3.2 '

Table 3.2.5 Reactor Protection System (Scram)

Instrumentation Requirements Scram Instrumentation Functional Tests Scram Instrumentation Calibration Instrumentation that Initiates Primary Containment Isolation Functions Instrumentation that Initiates or Controls the Core and Containment Cooling System Instrumentation that Initiates Rod Block

'adioactive Liquid Effluent Monitoring Instrumentation Radioactive Gaseous Effluent Monitoring Instrumentation 3/4.1 5

3/4. 1 8 3/4.1 -10 3/4.2 8

3/4.2 -10 3/4.2 -12 3/4.2 -14 3/4.2 -15 Table 4.2.1, Table 4.2.2 Table 4.2.3 Table 4.6.2 Table 3.7.1 Table 4.8.1 Table 4.8.2 Table 4.8.3 Table 4.8.4 Table 4.8.5 Table 4.8,6 Table Table Table Table Table Table 3.12-1 3.12-2 3.12-3 3.12-4 6.1.1 6.6 '

Table 4.11-1 Minimum Test and Calibration Frequency for Core and Containment Cooling Systems Instrumentation, Rod Blocks, and Isolations Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Neutron Flux and Sample Withdrawal Primary Containment Isolation Radioactive Gaseous Haste Sampling and Analysis Program Maximum Permissible Concentration of Dissolved or Entrained Noble Gases Released from the Site to Unrestricted Areas in Liquid Haste Radioactive Liquid Haste Sampling and Analysis Program Radiological Environmental Monitoring Program Reporting Levels for Radioactivity Concentrations in Environmental Samples Practical Lower Limits of Detection (LLD) for Standard Radiological Environmental Monitoring Program Surveillance Requirements for High Energy Piping Outside Containment Fire Detection Instruments Sprinkler Systems C02 Systems Fire Hose Stations Minimum Shift Manning Chart Special Reports 3/4.2 -17 3/4.2 -20.

3/4.2 -22 B 3/4.6 -26 3/4.7 -31 3/4.8 -22 3/4.8 -24 3/4.8 -25 3/4.8 -27 3/4.8 -28 3/4.8 -29 3/4.11-3 B 3/4.12-17 B 3/4.12-18 B 3/4.12-19 B 3/4.12-20 6 21 6

5 6

-'26 vli 3958a 0009A

I

~

y

~ r DRE III DPR-25 Amendment No. P,'P

, 78 List of Fi ures

~Pa e

Figure 3.5-1 Figur e 3. 5-1 Figure 3.5-1 Figure 3.5-1 Figure 3.5-1 Figure 3.5-2 Figure 3.6.1 Figure Figure Figure Figure Figure Figure 4.6.1 4,6.2 4.8-1 4.8-2 6.1-1 6.1-2 Figure 2.1-3 Figure 4.1.1 Figure 4.2.2 Figure 3.4.1 Figure 3.4.2 APRM Bias Scram Relationship to Normal Operating Conditions Graphical Aid in the Selection of an Adequate Interval Between Tests Test Interval vs.

System Unavailability Standby Liquid Control Solution Requirements Sodium Pentaborate Solution Temperature Requirements Bundle Average Exposure (MN)/MT)

(Sheet 1 of 5)

Planar Average Exposure (MN)/T)

(Sheet 2 of 5)

Planar Average Exposure (MN)/T)

(Sheet 3 of 5)

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)

(Sheet 4 of 5)

Planar Average Exposure (MWD/ST)

(Sheet 5 of 5)

Core Flow 4 Minimum Temperature Requirements per Appendix G of 10 CFR 50 Minimum Reactor Pressurization Temperature Chloride Stress Corrosion Test Results at 500 F

Owner Controlled/Unrestricted Area Boundary Detail of Central Complex Corporate Organization Station Organization B 1/2.1-17 B 3/4.1-18 B 3/4. 2-33 3/4.4-4 3/4.4-5 3/4.5-17 3/4.5-18 3/4.5-19 3/4.5-20 3/4.5-21 3/4.5-25

6. 26 3/4,6-20 B 3/4.6-25 B 3/4.6-27 B 3/4.8-38 B 3/4.8-39 6

3 6

4 viii 3958a 0009A

DRESDEN III DPR-25 Amendment No. g> ?8 3.6 LIMITING CONDITION FOR OPERATION (Cont'd.)

4.6 SURVEILLANCE RE UIREMENT (Cont'd.)

'I 2.

whenever both recirculation pumps are in operation, pump speeds shall be maintained within 107 of each other when power level is greater than 80% and within 157 of each other when power level i.s less than 80%.

If specification 3.6.H.l cannot be met, one recirculation pump shall be tripped.

Recirculation pumps speed shall be checked daily for mismatch.

3.

The reactor shall not be operat'ed with one recirculation loop out of service for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

With the reactor operating, if one recirculation loop is out of service the plant shall be placed in a hot shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless the loop is sooner returned to service.

4, Whenever one pump is operable and the remaining, pump is in the tripped position, the operable pump shall be at a speed less than 65% before starting the inoperable pump.

I.

Snubbers (Shock Suppressors)

I.

Snubbers (Shock)

Suppressors)

The following surveillance requirements apply to safety related snubbers.

3897a 3123A 3/4.6-12

DRESDEN III DPR-25 Amendment No. P.

78 3.6 LIMITING CONDITION FOR OPERATION (Cont'd,)

4.6 SURVEILLANCE RE UIREMENT (Cont'd.)

1.

During all modes of operation except cold shutdown and refuel, all safety related snubbers shall be operable except as noted in Specification 3.6.I.2 through 3 '.I.4.

1.

Visual Inspection An independent visual inspection shall be performed on the safety related hydraulic and mechanical snubbers in accordance with the schedule below.

a.

All hydraulic snubbers whose seal material has been demonstrated by operating experience, lab testing or analysis to be compatible with the operating environment shall be visually inspected.

This inspection shall

include, but not necessarily be limited to, inspection of the hydraulic fluid reservoir, fluid connections, and linkage connection to the piping and anchor to verify snubber operability.

b.

All mechanical snubbers shall be visually xnspected.

Thxs 3/4.6-13 3897a 3123A

DRESDEN III DPR-25 hmendment No.

7P

~ "8 3.6 LIMITING CONDITION FOR OPERATION (Cont'd.)

4.6 SURVEILLANCE RE UIREMENT (Cont'd.)

2.

From and after the time a snubber is determined to be inoperable, continued reactor operation is permissible only during the succeeding 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months unless the snubber is sooner made operable or replaced.

2.

Functional Testing a.

Once each refuel-ing cycle, a

representative sample of approxi-mately 107 of the hydraulic snubbers shall be function-ally tested for operability, includ-ing:

(i)

Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression, Snubber bleed, or release

rate, where required, is within the specified range in compression or tension.

3897a 3123h 3/4.6-15

DRESDEN III DPR-25 amendment No. $

7 78 3.6 LIMITING CONDITION FOR OPERATION (Cont'd.)

4.6 SURVEILLANCE RE UIREMENT (Cont'd.)

For each unit and subsequent unit found inoperable,.-an additional 107 of the hydraulic snubbers shall be tested until no more failures are found or all units have been tested.

b.

Once each refueling

cycle, a

representative sample of approximately 107 of the mechanical snubbers shall be functionally tested for operability.

The test shall consist of two parts:

Verification that the force that initiates free movement of the snubber in either tension or compression is less than the specified maximum breakaway friction force.

3/4.6-16 3897a 3123A

DRESDEN III N'R-25 Amendment No.

7+

~ <<

3 '

LIMITING CONDITION FOR OPERATION (Cont'd

~ )

4.6 SURVEILLANCE RE UIREMENT (Cont'd.)

)

Verify that the actxvation (restraining action) is achieved wxthxn the specified range of acceleration or

velocity, as applicable based on snubber design in both tension and compression.

For each unit and subsequent unit found inoperable, an additional 101.

of the mechanical snubbers shall be so tested until no more failures are found or all units have been tested.

c.

In addition to the regular

sample, snubbers which failed the previous functional test shall be retested during the next test period. If a spare snubber has been installed in place of a failed

snubber, then both the failed snubber (if it is repaired and installed in another position) and the spare snubber shall be

retested, Test results of these snubbers may not be included for the resampling

~

3897a 3123A 3/4.6-17

II I I

DRESDEN III NPR-25 Amendment No.

p7 3.6 LIMITING CONDITION FOR OPERATION (Cont'd.)

4.6 SURVEILLANCE RE UIREMENT (Cont'd.)

3

~ If the requirements of 3'.I,l and 3.6.I.2 cannot be

met, an orderly shutdown shall be initiated and the reactor shall be in cold shutdown or refuel condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

3.

Mhen a snubber is deemed inoperable, a review of all pertinent facts shall be conducted to determine the snubber mode of failure and to decide if an engineering evaluation should be performed on the supported system or components.

If said evaluation is deemed necessary, it will determine whether 6r not the snubber mode of failure has imparted a

significant effect or degradati,on on the supported component or system.

4. If a snubber is determined to be inoperable while the reactor is in the cold shutdown or refuel mode, the snubber shall be made operable or replaced prior to reactor startup.

4. If any snubber selected for functional testing either fails to lock up or fails to move, i.e., frozen in

place, the cause will be evaluated and, if determined to be a

generic deficiency, all snubbers of the same design subject to the same defect shall be functionally tested.

3897a 3123A 3/4.6-18

DRESDEN III DPR-25 Amendment No. g

~

78 3.6 LIMITING CONDITION FOR OPERATION (Cont'd.)

4.6 SURVEILLANCE RE UIREHENT (Cont'd.)

5.

Snubbers may be added or removed from safety related systems without prior license amendment.

5.

Snubber service life monitoring shell be followed by existing station record systems, including the central filing system, maintenance files, safety related work

packages, and snubber inspection records.

The above record retention methods shall be used to prevent the hydraulic snubbers from exceeding a service life of 10 years and the mechanical snubbers from exceeding a

service life of 40 years (lifetime of the plant).

3897a 3123A 3/4.6-19

DRESDEN III

~R-25 Amendment No. P e

iSnirnwa Teeaperawre Requirement per Appendix G ot 10 CF A 50 1100 1000 900 Z

eoo 200 er~.

600 5

5oo

~.

400 r

300 CURVE,A IINSEAVICE PAESSUAE TESTS-SECTION XI1 I

I CUAVE B IHEATUP ~ COOLOOWN)

CURVE C {CRITICALCOAE OPEAATIONI MINIMUMBOLTING TEMPERATURE 100 F

MINIMUMOPERATING TEMPERATURE" 149 F

RTNOT 40 F

Ki PER SECTION G 21IO OF APPENDIX G OF THE SUMMER 1973 AOOENOA TO SECTION III OF THE ASME COOE TEMPERATURE

( F)

Fig. 3.6.1 MINIMUM TEMPERATURE REQUIREMENTS PER APPENDIX G OF 10 CFR 50 3897a 3123A 3/4.6-20

DRESDEN III QPR-25 Amendment No.

3.6 LIMITING CONDITION FOR OPERATION BASES h.

Thermal Limitations The reactor vessel design specification requires that the reactor vessel be designed for a maximum heatup and cooldown rate of the contained fluid (water) of 100 F per hour averaged over a period of one hour.

This rate has been chosen based on past experience with operating power plants.

The associated time periods for heatup and cooldown cycles when the 100'F per hour rate is limiting provides for efficient, but safe, plant operation.

The reactor vessel manufacturer has designed the vessel to the above temperature criterion.

In the course of completing the

~

design, the manufacturer performed detailed stress

analysis, This analysis includes more severe thermal conditions than those which would be encountered during normal heating and cooling operations.

Specific analyses were madebased on a heating and cooling rate of 100 F/hour applied continuously over a temperature range of 100 F to 550'F.

Because of the slow temperature-time response of the massive flanges relative to the adjacent head and shell sections, calculated temperatures obtained were 500'F (shell) and 360 F (flange)

(140 F differential).

Both axial and radial thermal stresses were considered to act concurrently with full primary loadings.

Calculated stresses were within ASME Boiler and Pressure Vessel Code Section III stress intensity and fatigue limits even at the flange area where maximum stress occurs.

The flange metal temperature differential of 140 F occurred as a result of sluggish temperature response and the fact that the heating rate continued over a 450 F coolant temperature range.

The uncontrolled cooldown rate of 240'F was based on the maximum expected transient over the lifetime of the reactor vessel.

This maximum expected transient is the injection of cold water into the vessel by the high pressure coolant injection subsystem.

This transient was considered in the design of the pressure vessel and five such cycles were considered in the design.

Detailed stress analyses were conducted to assure that the injection of cold water into the vessel by the HPCI would not exceed ASME stress code limitations.

B.

Specification 3.6.h.4 increases margin of safety for thermal-hydraulic stability and startup of recirculation pump.

3897a 3123h B 3/4,6-21

DRESDEN III QPR-25 Amendment No. g

~

3,6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

Pressurization Tem erature The reactor coolant system is a

primary barrier against the release of fission products to the environs.

In order to provide assurance that this barrier is maintained at a high degree of integrity, restrictions have been placed on the operating conditions to which it can be subjected.

These restrictions on inservice hydrostatic testing on heatup and cooldown, and on critical core operation shown in Figure 3.6.1, were established to be in conformance with Appendix G to 10 CFR 50.

In evaluating the adequacy of ferritic steels Sa302B it is necessary that the following be established:

a)

The reference nil-ductility temperature (RTNDT) for all vessel and adjoining materials, b) the relationship between RTNDT and integrated neutron flux (fluence, at energies greater than one Mev), and c) the fluence at the location of a postulated flow.

The initial RTNDT of the main closure flange, the shell and head materials connecting to these flanges, and connecting welds is 10 F.

However, the vertical electroslag welds which terminate iaaaediately below the vessel flange have an RTNDT of 40'F.

(Reference Appendix F to the FSAR)

The closure flanges and connecting shell materials are not subject to any appreciable neutron radiation exposure, nor are the vertical electroslag seams.

The flange area is moderately stressed by tensioning the head bolts.

Therefore, as is indicated in curves (a) ind (b) of Figure 3.6.1, the minimum temperature of the vessel shell imnediately below the vessel flange is established as 100'F below a pressure of 400 psig.

(40 F +

60'F, where 40 F is the RTNDT of the electroslag weld and 60 F is a conservatism required by the ASME Code).

Above approximately 400 psig pressure, the stresses associated with pressurization are more limiting than the bolting stresses, a

fact that is reflected in the non-linear portion of curves (a) and (b).

Curve (c), which defines the temperature limitations for critical core operation, was established per Section IV 2.c. of Appendix G of 10CFR50.

Each of the curves, (a),

(b) and (c) define temperature limitations for unirradiated ferric steels.

Provision has been made for the modification of these curves to account for the change in RTNDT as a result of neutron embrittlement, 3897a 3123A B 3/4.6-22

DRESDEN III DPR-25 Amendment No. g~

78 3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

The withdrawal schedule in Table 4.6.2 is based on the three capsule surveillance program as defined in Section 11.C.3.a of 10 CFR 50 Appendix H.

The accelerated capsule (Near Core Top Guide) is not required by Appendix H but will be tested to provide additional information on the vessel materi:al.

This surveillance program conforms to ASTM E 185-73 "Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels" with one exception.

The base metal specimens of the vessel were made with their longitudinal axes parallel to the principal rolling direction of the vessel plate.

C.

Coolant Chemistr

- h radioactivity concentration limit of 20 Micro-Ci/ml total iodine can be reached if the gaseous effluents are near the limit as set forth in Specification 3.8.C.l or there is a failure or a prolonged shutdown of the cleanup demineralizer.

In the event of a steam line rupture, outside the drywell, the resultant radiological dose at the site boundary would be about 10 rem to the thyroid.

This dose was calculated on the basis of a total iodine activity limit of 20 Micro-Ci/ml, meteorology corresponding to Type F

conditions with a one meter per second wind speed, and a valve closure time of five seconds.

If the valve closed in ten

seconds, then the resultant dose would increase to about 25 rem.

The reactor water sample will be used to assure that the limit of Specification 3.6.C is not exceeded.

The total radioactive iodine activity would not be expected to change rapidly over a

period of 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months .

In addition, the trend of the stack off-gas release

rate, which is continuously monitored, is a

good indicator of the trend of the iodine activity in the reactor coolant.

Since the concentration of radioactivity in the reactor coolant is not continuously measured, coolant sampling would be ineffective as a means to rapidly detect gross fuel element failures.

However, some capability to detect gross fuel element failures is inherent in the radiation monitors in the off-gas system and on the main steam lines.

Materials in the primary system are primarily 304 stainless steel and the Zircaloy fuel cladding.

The reactor water chemistry limits are established to prevent damage to these materials.

Limits are placed on chloride concentration and conductivity.

The most important limit is that placed on chloride concentration to prevent stress corrosion cracking of 3897a 3123h 8 3/4.6-23

P

(

I )

DRESDEN III DPR-25 Amendment No. 7'~

3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

the stainless steel.

The attached graph, Figure 4.6.2, illustrates the results of tests on stressed 304 stainless steel specimens.

Failures occurred at concentrations above the curve; no failures occurred at concentrations below the curve.

According to the data, allowable chloride concentrations could be set several orders of magnitude above the established limit, at the oxygen concentration (0.2-0.3 ppm) experienced during power operation.

Zircaloy does not exhibit similar stress corrosion failures.

However, there are various conditions under which the dissolved oxygen content of the reactor coolant water could be higher than 0.2-0.3

ppm, such as refueling, reactor startup and hot standby.

During these periods with steaming rates less than 100,000 pounds per hour, a more restrictive limit of 0.1 ppm has been established to assure the chloride-oxygen combinations of Figure 4.6.2 are not exceeded.

At steaming rates of at least 100,000 pounds per hour, boiling occurs causing deaeration of the reactor water, thus maintaining oxygen concentration at low levels.

When conductivity is in its proper normal range, pH and chloride and other impurities affecting conductivity must also be within their normal range.

When and if conductivity becomes

abnormal, then chloride measurements are made to determine whether or not they are also out of their normal operati,ng values.

This would not necessarily be the case.

Conductivity could be high due to the presence of a neutral salt; e.g.,

Na2S04, which would hot have an effect on pH or chloride.

In such a case, high conductivity alone is not a

cause for shutdown.

In some types of water;cooled reactors, conductivities are in fact high due to purposeful addition of additives.

In the case of BWR's, however, where no additives are used and where neutral pH is maintained, conductivity provides a very good measure of the quality of the reactor water.

Significant changes therein provide the operator with a warning mechanism so he can investigate and remedy the condition causing the change before limiting conditions, with respect to variables affecting the boundaries of the reactor

coolant, are exceeded.

Methods available to the operator for correcting the off-standard condition include, operation of the reactor clean-up system,, reducing the input of impurities and placing the reactor in the cold shutdown condition.

The major benefit of cold shutdown is to reduce the temperature dependent corrosion rates and provide time for the clean-up system to re-establish the purity of the reactor coolant.

3897a 3123A B 3/4.6-24

DRESDEN III DPR-25 Amendment No. p s

78 I

R l?0 TYPlCALCURVE 40 Sol S loll NTEChATEO IKMThOK CXPOSlhlC > l WV ~u lO" Figure 4,6.1 MINIMUM REACTOR PRESSURIZATION TEMPERATURE 3897a 3123A B 3/4.6-25

DRESDEN III DPR-25 hmendment No.

TABLE 4.6.2 NEUTRON FLUX hND ShMPLES WITHDRAWAL SCHEDULE FOR DRESDEN UNIT 3 Withdrawal Year Part No.

Location Comments 1978 16 Near Core Top Guide - 180'ccelerated 1981 18 Mall 215 2001 19 15 20 Wall 245 Mall 65 Wall 275 Standby Standby 3897a 3123h B 3/4.6-26

DRESDEN III DPR-25 Amendment No. p, 78 100 fAILURE

~.I Llml Ealabluhcd los CI st OPersling O> Cmesusws

= 03 -O.d pps RO fAILURE

~.01 0.1

1.0 REfEREIICE

"CORROQOR 0 Wf*h HAIIOOOOK D. J. Oefavl, Ed.

10 CNLORIOE ~I 100 Figure 4.6.2 CHLORIDE STRESS CORROSION TEST RESULTS AT 5000F 3897a 3123A B 3/4.6-27

DRESDEN III DPR-25 Amendment No. g, 78 3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

During start'-up periods, which are in the category of less than 100,000 pounds per hour, conductivity may exceed 2

micro-mho/cm because of the initial evolution of gases and the initial addition of dissolved metals.

During this period of

time, when the conductivity exceeds 2 micro-mho (other than short term spikes),

samples will be taken to assure the chloride concentration is less than 0.1 ppm.

The conductivity of the reactor coolant is continuously monitored.

The samples of the coolant which are taken every 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months will serve as a reference for calibration of these monitors and is considered adequate to assure accurate readings of the monitors. If conductivity is within its normal range, chlorides and other impurities will also be within their normal ranges.

The reactor coolant samples will also be used to determine the chlorides.

Therefore, the sampling frequency is considered adequate to detect long-term changes in the chloride ion content.

Isotopic analyses required by Specification 4.6.C.3 may be performed by a gamma scan.

D.

Coolant Leaka e

Allowable leakage rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes and on the ability to makeup coolant system leakage in the event of loss of offsite a-c power.

The normally expected background leakage due to equipment design and the detection capability for determining coolant system leakage were also considered in establishing the limits.

The behavior of cracks in piping systems has been experimentally and analytically investigated as part of the USAEC sponsored Reactor Primary Coolant System Rupture Study (the Pipe Rupture Study).

Work utilizing the data obtained in this study indicates that leakage from a crack can be detected before the crack grows to a dangerous or critical size by mechanically or thermally induced cyclic loading, or stress corrosion cracking or some other mechanism characterized by gradual crack growth.

This evidence suggests that for leakage somewhat greater than the limit specified for unidentified leakage, the probability is small that imperfections or cracks associated with such leakage would grow rapidly.

However, the establishment of allowable unidentified leakage greater than that given in 3.6.D on the basis of the data presently available would be premature because of uncertainties associated with the data.

For leakage of the order of 5 gpm as specified in 3.6.D, the experimental and analytical data suggest a reasonable margin of safety that such leakage magnitude would not result from a crack approaching the critical size for rapid propagation.

Leakage less than the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available 3891a 3123A B 3/4.6-28

DRESDEN III 3)PR-25 Amendment No. g

~ /8 3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

leakage detection

schemes, and if the origin cannot be determined in a reasonably short time the plant should be shut down to allow further investigation and corrective action.

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

Removal of 50 gpm from either of these sumps can be accomplished with considerable margin.

The performance of reactor coolant leakage detection system will be evaluated during the first five years of station operation and the conclusions of this evaluation will be reported to the NRC.

It is estimated that the main steam line tunnel leakage detection system is capable of detecting the order of 3000 lb/hr.

The system performance will be evaluated during the first five years of plant operation and the conclusions of the evaluation will be reported to the NRC.

ED Safet and Relief Valves - The frequency and testing requirements for the safety and relief valves are specified in the Inservice Testing Program which is based on Section RI of the ASME Boiler and Pressure Vessel Code.

Adherence to these code requirements provides adequate assurance as to the proper operational readiness of these valves.

The tolerance value is specified in Section III of the ASME Boiler and Pressure Vessel Code as plus or minus 1% of design pressure.

An analysis has been performed which shows that with all safety valves set 1f higher than the reactor coolant pressure safety limit of 1375 psig is not exceeded.

The safety valves are required to be operable above the design pressure (90 psig) at which the core spray subsystems are not designed to deliver full flow.

F.

Str ctural inte rit A pre-service inspection of the components in the primary coolant pressure boundary will be conducted after site erection to assure the system is free of gross defects and as a reference base for later inspections.

Prior to operation, the reactor primary system will be free of gross defects.

In addition, the facility has been designed such that gross defects should not occur throughout life.

Inservice Inspections of ASME Code Class 1,

2 and 3 components will be performed in accordance with the applicable version of Section RI of the ASME Boiler and Pressure Vessel Code.

3897a 3123h B 3/4,6-29

DRESDEN III 7gPR-25 Amendment No. Q

'.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

Relief from any of the above requirements must be provided in writing by the Commission.

The Inservice Inspection program and the written relief do not form a part of these Technical Specifications.

. These studies show that it requires thousands of stress cycles at stresses beyond any expected to occur in a reactor system to propagate a crack.

The test frequency established is at intervals such that in comparison to study results only a small number of stress

cycles, at values below limits will occur.

On this basis, it is considered that the test frequencies are adequate.

The type of inspection planned for each component depends on location, accessibility, and type of expected defect.

Direct visual examination is proposed wherever possible since it is sensitive, fast and reliable.

Magnetic particle and liquid penetrant inspections are planned where practical, and where added sensitivity is required.

Ultrasonic testing and radiography shall be used where defects can occur on concealed surfaces.

After five years of operation, a program for in-service inspection of piping and components within the primary pressure boundary which are outside the downstream containment isolation valve shall be submitted to the NRC.

G.

~Jet Pum s

Pailuse of a jet pump nozzle assembly hold doom mechanism, nozzle assembly and/or riser increases the cross sectional flow area for blowdown following the postulated design basis double-ended recirculation line break.

Therefore, if a failure occurs, repairs must be made to assure the validity of the calculated consequences.

The following factors form the basis for the surveillance requirements:

h break in a jet pump decreases the flow resistance characteristic of the external piping loop causing the recirculation pump to operate at a higher flow condition when compared to previous operation.

The change in flow rate of the failed jet pump produces a

change in the indicated flow rate of that pump relative to the other pumps in that loop.

Comparison of the data with a normal relationship or pattern provides the indication necessary to detect a failed jet pump, 3897a 3123A B 3/I4.6-30

DRESDEN III DPR-25 Awent No. $5',

78 3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd,)

The jet pump flow deviation pattern derived from the diffuser to lower plenum differential pressure readings will be used to further evaluate jet pump operability in the event that the jet pumps fail the tests in Section 4.6.G.1 and 2.

Agreement of indicated core flow with established power-core flow relationships provides the most assurance that recirculation flow is not bypassing the core through inactive or broken jet pumps.

This bypass flow is reverse with respect to normal jet pump flow, The indicated total core flow is a

sunmation of the flow indications for the twenty individual jet pumps.

The total core flow measuring instrumentation sums reverse jet pump flow as though it were forward flow.

Thus the indicated flow is higher than actual core flow by at least twice the'ormal flow through any backflowing pump.

Reactivity inventory is known to a high degree of confidence so that even if a jet pump failure occurred during a shutdown

period, subsequent power ascension would promptly demonstrate abnormal control rod withdrawal for any power-flow operatng map point.

h nozzle-riser system failure could also generate the coincident failure of a jet pump body; however, the 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.

H.

Recirculation Pum Flow Mismatch The LPCI loop selection logic has been described in the Dresden Nuclear Power Station Units 2 and 3 FSAR, Amendments 7

and 8.

For some limited low probability accidents with the recirculation loop operating with large speed differences, it is possible for the logic to select the wrong loop for injection.

For these limited conditions, the core spray itself is adequate to prevent fuel temperatures from exceeding allowable limits.

However, to limit the probability even

further, a procedural limitation has been placed on the allowable variation in speed between the recirculation pumps.

The licensee's analyses indicate that above 80% power the loop select logic could not be expected to function at a speed differential of 15%.

Below 8M power, the loop select logic would not be expected to function at a speed differential of 207.

This specification provides a margin of 5% in pump speed differential before a problem could arise.

If the reactor is operating, on one pump, the loop select logic trips that pump before making the loop selection.

3897a 3123A B 3/4.6-31

DRESDEN III QPR-25 Amendment No.

3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

In addition, during the start-up of Dresden Unit 2, it was found that a flow mismatch between the two sets of jet pumps caused by a difference in recirculation loops could set up a

vibration until a mismatch in speed of 27% occurred.

The 101.

and 15'peed mismatch restrictions provide additional margin before a

pump vibration problem will occur.

ECCS performance during reactor operation with one recirculation loop out of service has not been analyzed.

Therefore, sustained reactor operation under such conditions is not permitted.

I; Snubbers Shock Su ressors)

Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient while allowing normal thermal motion during startup and shutdown.

The consequence of an inoperable-snubber is an increase in the probabi.lity of structural damage to piping as a result of a seismic or other event initiating dynamic loads.

It is therefore required that all snubbers required to protect the primary coolant system or any other safety system or component be operable during reactor operation.

Because the snubber protection is required only during low probability events, a period of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed for repairs or replacements.

In case a shutdown is required, the allowance of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 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 equipment, Specification 3.6.I.4 prohibits startup with inoperable snubbers.

When a snubber is found inoperable, a review shall be performed to determine the snubber mode of failure.

Results of the review shall be used to determine if an engineering evaluation of the safety-related system or component is necessary.

The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on the support component or system.

All safety related hydraulic snubbers are visually inspected for overall integrity and operability.

The inspection will include verification of proper orientation, adequate hydraulic fluid level and proper attachment of snubber to piping and structures.

3897a 3123A B 3/4.6-32

I

~

Amendment No.

/7

~

3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

All safety related mechanical snubbers are visually inspected for overall integrity and operability.

The inspection will include verification of proper orientation and attachments to the piping and anchor for indication of damage or impaired operability.

, The inspection frequency is based upon maintaining, a constant level of snubber protection.

Thus, the required inspection interval varies inversely with the observed snubber failures.

The number of inoperable 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 inspection.

However, 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.

To further increase the assurance of snubber reliability, functional tests will be performed once each 'refueling cycle.

A representative sample of 107 of the safety-related snubbers will be functionally tested.

Observed failures on these samples will require testing of additional units'ydraulic snubbers and mechanical snubbers may each be treated as different entities for the above surveillance programs.

Hydraulic snubber testing, will include stroking of the snubbers to verify piston movement, lock-up, and bleed.

Functional testing of the mechanical snubbers will consist of verification that the force that initiates free movement of the snubber in either tension or compression is less than the maximum breakaway friction force and verification that the activation (restraining action) is achieved within the specified range of acceleration or velocity, as applicable based on snubber

design, in both tension and compression.

B 3/4.6-33 3897a 3123A

4 )

DRES~ III DPR-25 Ame~nt No. g~

78 3.6 LIMITING CONDITION FOR OPERATION BASES (Cont'd.)

When the cause of rejection of the snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible, and verified by inservice functional testing, that snubber may be exempted from being counted as inoperable.

Generically susceptible snubbers are those which are of a specific make or model and

, have the same design features directly related to rejection of the snubber by visual inspecti.on or are similarly located or exposed to the same environmental conditions such as temperature, radiation, and vibration.

Monitoring of snubber service life shall consist of the existing station record systems, including the central filing

system, maintenance files, safety-related work packages, and snubber inspection records.

The record retention programs employed at the station shall allow station personnel to maintain snubber integrity.

The service life for hydraulic snubbers is 10 years.

The hydraulic snubbers existing locations do not impose undue safety implications on the piping and components because they are not exposed to excesses in environmental conditions.

The service life for mechanical snubbers is 40 years, lifetime of the plant.

The mechanical snubbers are installed in areas of harsh environmental conditions because of their dependability over hydraulic snubbers in these areas.

All snubber installations have been thoroughly engineered providing the necessary safety requirements.

Evaluations of all snubber locations and environmental conditions justify the above conservative snubber service lives.

4.6 SURVEILLANCE RE UIREMENT BASES None B 3/4.6-34 3897a 3123A

( p I

4

5KZZZNR%9gSTR'giE COPE

~~ >6i 1985

@)Qg $~og~

Docket Nos.

Q-250 and 50-51 By letter dated February 1,

1985, you indicate that Westinghouse informed you, subsequent to your September 28, 1984 letter, that administrative controls on fuel loading are required for racks whose outer rows overhang the support pads in order to be consistent with an assumption by

. Westinghouse during its analysis.

That is, the outer (overhanging) rows would not be fully loaded while the remaining portion of the rack module is empty.

The NRC staff's SE and the supporting TER conclusions have remained valid due to the administrative controls initiated when you became aware of the potential need for the controls.

These controls, which were prior to any fuel loading in the affected racks, preclude the possibility of any li t-off.

, Your February 1,

1985, letter requested that we review the information provided as the result of a reanalysis of fuel racks with only overhanging rows loaded with fuel which indicates the worse case lift-offis less than 0.2 i'nches during a seismic event and this minimal lift-offwilT not result in failures to the racks or pool structures and their integrity will be maintained regardless of the loading pattern.

Distribution Docket file ORB¹l RDG Gray file (4)

HThompson CParrish Mr. J.

W. Williams, Jr., Vice President DMcDonald EJordan Nuclear Energy Department PMcKee OELD Florida Power and Light Company ACRS (10)

DBrinkman Post Office Box 14000 TBarnhart (8)

'JPartlow Juno

Beach, Florida 33408 WJones A, CMiles RDiggs

Dear Mr. Williams:

Bases Weve~

Reference:

TAC Nos.

56805 and 56808

SUBJECT:

SPENT FUEL STORAGE FACILITY EXPANSION l(l

OPR.-al By letter dated November 21, 1984, the Commission issued Amendment No.

111 to Facility Operating License No.

DPR-31 and Amendment No.

105 to Facility Operating License No.

DPR-41 for the Turkey Point Plant Units 3 and 4, respectively, which allowed expansion of the spent fuel storage facilities.

Copies of the supporting Safety Evaluation and Notice of Issuance and Final Determination of No Significant Hazards Consideration were also enclosed.

The Safety Evaluation (SE) and the appended Technical

~ Evaluation Report (TER) provided the basis for our issuance of the requested amendments.

Sections 2.3.4 and 2.3.5 of the SE and the appended TER indicated that postulated loads from a seismic event will not result in failures to the racks or pool structures, thus their integrity will be maintained.

As indicated in Section 3.3.4 of the TER, there would be no lift-offof the rack modules from the pool liner during a seismic event.

This conclusion was based on your September 28, 1984, letter which provided the results of the Westinghouse analysis.

~ MXK xMxzxmamazy

~

'l

'8(

0 Mr. Williams February 26, 1985 This request for our review of the reanalysis represents a chan e in a basis su ortin the above referenced amendments as documented in the supporting afety va ua ion.

R 5%59, "Changes tests and experimentP" indicates that licensee's may make changes, conduct tests or experiments not described in the Safety Analysis Report without prior Commission approval unless the proposed

change, test or experiment involves a change in the technical specifications incorporated in the license or an unreviewed safety question.

It is not clear from your submittal whether you have performed a 50.59 review and documented the results in accordance with the provisions of 50.59(a) and (b) or; that you have determined that the reanalysis requires a change in the technical specifications incorporated in the licenses or that the change represents an unreviewed safety question.

'f you have performed a 50.59 review in accordance with the provisions of 50.59(a) and.(b) and determined that neither an explicit technical specification change hor an unreviewed safety question is involved, you do not need our prior approval and your request may be withdrawn.

However, if

. you have determined.

a change in the technical specifications incorporated in the license or an unreviewed safety question exists, we request that your submittal be modified in accordance with 50.59(c) including a proposed Notice for public comment using the standards in 10 CFR 50.92 concerning the issue of no significant hazards consideration.

We will take no further action on this request until we receive clarification.

The reporting and/or recordkeeping requirements of this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.

Sincerely,

/s/ Omcoonala Daniel G. McDonald, Project Manager Operating Reactors Branch 5'1 Division of Licensing ee previous white for concurrences ORB81: DL*

CParrish 02/ 26/85

'DL DMcDonal d/ts 02/y@85 BC-ORBPl: DL*

SVarga 02/26/85 OELD*

02/ 26/85=

0

This request for our review of the reanalysis represents a change in the design basis for the above referenced amendments as documented fn the'upporting Safety Evaluation.

10 CFR 50.59, "Changes tests and experiments,"

indicates that licensee's may make changes, conduct tests or experiments not described in the Safety Analysis Report without prior Commission approval unless the proposed change, test or experiment involves a change in the technical specifications incorporated in the license or an unreviewed safety question.

It is not clear from your submittal whether you have performed a 50.59 review and documented the results in accordance with the provisions of 50.59(a) and (b) or; that you have determined that the reanalysis requires a change in the technical, specifications incorporated in the licenses or that the change represents an unreviewed safety question.

If you have performed a 50.59 review in accordance with the provisions of 50.59(a) and (b), you do not need our prior approval and your request may be withdrawn.

However, if you have determined a change in the technical specifications incorporated. in the license or an unreviewed safety question

exists, we request that your submittal be modified in accordance with 50.59(c) including a proposed Notice for public comment using the standards in 10 CFR 50.92 concerning the issue of no significant hazards consideration.

We will take no further action on this request until we receive clarification.

The reporting and/or recordkeeping requirements of this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.

Sincerely, Enclo e:

tated Daniel G. McDonald, Project Manager Operating Reactors Branch ¹1 Division of Licensing CC See next page ORB¹1:DL CParrish 02/g/85 ORB¹1:DL BC-0

¹1: DL DMcDonald/ts SV~

02/g4'85 02@4/85 02;0 /8

J;>W. Williams, Jr.

Florida Power and Light Company Turkey Point Plants Units 3 and 4

CC:

Harold F. Reis, Equire Newman and Holtzinger; P.C.

1615 L Street, N.W.

Washington, DC 20036 Mr. Jack Shreve Office of the Publ ic Counsel Room 4, Holland Building Tallahassee, Florida 32304 Norman A. Coll, Esquire

~ Steel, Hector and Davis 4000 Southeast Financial Center Miami, Florida 33131-2398 Mr.

Ken N. Harris, Vice President Turkey Point Nuclear Plant Florida Power and Light Company P.0.

Box 029100 Miami, Florida 33102 Mr.

M.

R. Stierheim County Manager of Metropolitan Dade County Miami, Florida 33130 Resident Inspector Turkey Point Nuclear Generating Station U.S. Nuclear Regulatory Commission Post Office Box 57-1185 Miami, Fl orida 33257-1185 Regional Radiation Representative EPA Region IV 345 Courtland Street, N.W.

Atlanta, GA 30308 Intergovernmental Coordination and Review Office of Planning 8 Budget Executive Office of the Governor The Capitol Building Tal'lahassee, Florida 3)301 Administrator-Department of Environmental Regulation Power Plant Siting Section State of Florida 2600 Blair Stone Road Tallahassee, Florida 32301 James P. O'Reilly Regional Administrator, Region II U.S Nuclear Regulatory Commission Suite 2900 101 Mari'etta Street

Atlanta, GA 30303 Martin H. Hodder, Esquire 1131 N.E. 86th Street Miami, Florida 33138 Joette Lorion 7269 SW 54 Avenue Miami, Florida 33143 Mr. Chris J.

Baker, Plant Manager Turkey Point Nuclear Plant Florida Power and Light Company P.O.

Box 029100 Miami, Florida 33102 Attorney General Department of Legal Affairs The Capitol Tallahassee, Florida 32304 Mr. Allan Schubert, Manager Public Health Physicist Department of Health and Rehabilitative Services 1323 Winewood Blvd.

Tallahassee, Florida 32301

J S

ML17346B341

Text

Attachment:

Attachment:

1.0 REfEREIICE

Dear Mr. Williams:

Reference:

SUBJECT:

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