ML20140C830

Amend 93 to License DPR-35,removing Details of ASME Boiler & Pressure Vessel Code Section XI Inservice Insp Program & Snubber Tables from Tech Specs & Requiring Placement in Facility Controlled Documents

ML20140C830
Person / Time
Site:	Pilgrim
Issue date:	03/17/1986
From:	Zwolinski J Office of Nuclear Reactor Regulation
To:
Shared Package
ML20140C834	List: ML20140C830 ML20140C850
References
NUDOCS 8603250501
Download: ML20140C830 (16)

Text

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pa u% '9[o UNITED STATES j,

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j BOSTON EDISON COMPANY DOCKET N0. 50-293 PILGRIM NUCLEAR POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE J

Amendment No. 93 License No. DPR-35 1.

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

A.

The application for amendment by Boston Edison Company (the licensee) dated April 17, 1985, as amended on September 24, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission'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 (1) that the activities authorized by this amendment can be conducted without endangering the health i

and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; i

D.

The issuance of this amendment will not be inimical to the.

comon 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 Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-35 is hereby 1

amended to read as follows:

8603250501 060317 PDR ADOCK 05000293 p

PDR

' B.

Technical Specifications The Technical Specifications contained in. Appendix A, as revised through Amendment No. 93, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the l

Technical Specifications.

3.

This license amendment is effective 30 days after the date of issuance.

FOR THE NUCLEAR REGULATO C0 ISSION

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John

. Zwolinski, Director

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Division of BWR Licensing BWR Pr ject Directorate #1 i

Attachment:

Changes to the Technical Specifications Date of Issuance: March 17,1986 1

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ATTACHMENT TO LICENSE AMENDMENT N0. 93 i

FACILITY OPERATING LICENSE NO. OPR-35 i

DOCKET NO. 50-293 i

l Revise the Appendix A Technical Specifications by' removing the pages identified below and inserting the attached pages. The revised pages are j

identified by the captioned amendment number and contain marginal lines indicating the area of change.

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REMOV'E INSERT i

i 127 127 i

127A 127A 1278 177B 129 137 130 137a 4

131 137b 132 137d 133 149 134 150 135 151 136 224 137 137a 137b i

137d i

137e i

137f j

1379 137h 1371 138A j

149 150 151 151a i

151b l

224 i

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

LIMITING CONDITIONS FOR OPERATION SURVEXLLANCE REQUIREMENTS l

3.6.D Safety Relief Valves (Con't)

E. Jet Pumps from the initial discovery of Whenever there is recirculation discharge pipe temperatures in flow with the reactor in the excess of 212*F for more than startup or run modes, jet pump 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without prior NRC operability shall be checked daily approval of the engineering by verifying that the following evaluation delineated in 3.6.D.3.

conditions do not occur

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simultaneously.

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The limiting conditions of operation for the instrumentation

1. The two recirculation loops have a that monitors tall pipe tempera-flow imbalance of 15% or more when 1

ture are given in Table 3.2.F.

the pumps are operated at the same speed.

E.

Jet Pumps J

2. The indicated value of core flow 1.

Whenever the reactor is in the rate varies from the value derived startup or run modes, all jet from loop flow measurements by more pumps shall be operable.

If it than 10%.

is determined that a jet pump is inoperable, an orderly shutdown

3. The diffuser to lower plenum shall be initiated and the differential pressure reading on an 4

reactor shall be in a Cold individual jet pump varies from Shutdown Condition within established jet pump P

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

characteristics by more than 10%.

F.

Jet Pump Flow Mismatch F. Jet Pump Flow Mismatch i

1.

Whenever both recirculation Recirculation pump speeds shall be pumps are in operation, pump checked and logged at least once

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i speeds shall be maintained within per day.

10% of each other when power level is greater than 80% and within 15% of each other when power level is less than or equal to 80%.

2.

If Specification 3.6.F.1 is exceeded immediate corrective action shall be taken.

If recirculation pump speed mismatch is not corrected within 30 minutes, an orderly shutdown shall be initiated and the reactor shall be in the Cold Shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the recirculation i

pump speed mismatch is brought within limits sooner.

e Amendment No. 71, 93 127'

LIMITING CONDITIONS FOR OPERATION SURVEZLLANCE REQUlREMENTS 1

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3.6.G Structural Integrity 4.6.G Structural Integrity j

1.

The structural integrity of Inservice inspection of j

the primary system boundary components shall be performed shall be maintained at the in accordance with the PNPS i

level required by the ASME Inservice Inspection Program.

Boller and Pressure Vessel The results obtained from Code,Section XI " Rules for compliance with this program Inservice Inspection of Nuclear will be evaluated at the Power Plant Components",

completion of each ten year 4

Articles IWA, IWB, IWC, IND and interval.

The conclusions of IWF and mandatory appendices as this evaluation will be reviewed j.

required by 10CFR50, Section with the NRC.

i 50.55a(g), except where i

specific relief has been i

granted by the NRC pursuant

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to 10CFR50, Section j

50.55a(g)(6)(1).

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l Amendment No.19, 93 127A I

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. LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.H High Energy Piping (outside 4.6.H High Energy Piping (outside

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containment) containment) 1.

The high energy line sections The inspections listed in

. identified in Table 4.6.2 shall Table 4.6.2 shall be performed be maintained free of visually as specified to verify the i

observable through-wall leaks.

structural integrity of the specified high energy line 4

2.

If a leak is detected by the sections.

The visual surveillance program of 4.6.H, inspection for leakage shall efforts to identify the source be consistent'with the

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of the leak shall be started requirements of ASME Boller and immediately.

Pressure Vessel Code,Section XI, 1980 Edition, Winter 1980 3.

If the source of leakage cannot Addenda, Subarticle INA-5240.

be identified within eight hours i

of detection or if the leak is l

found to be from the pressure j

retaining boundary in the

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sections identified in Table 4.6.2, the leak shall be isolated j

or the reactor shall be in a cold shutdown condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

i 4.

When the modifications, described in FSAR Amendment No. 34, to l

provide protection against high j

energy line breaks outside of the primary containment have been i

completed, Technical Specifications 3.6.H and 4.6.H will no longer be required.

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i Amendment No.19,93 127B l

PAGES 129 THROUGH 136 ARE DELETED Amendment No. 19, 93

l TABLE 4.6.2 INSPECTION REQUIREMENTS FOR HIGH ENERGY l

LINES OUTSIDE CONTAINMENT ITEM NO.

HIGH ENERGY AREA INSPECTION METHOD

• FREQUENCY 1.

Main steam lines outside Visual Monthly When containment from contain-Operating ment to turbine stop valves 2.

HPCI steam line in torus Visual Monthly When area and in HPCI turbine Operating area 3.

RCIC steam line in valve Visual Monthly When compartment and pump Operating compartment 4

RHCU line in pump, heat Visual Monthly When exchanger compartments and Operating valve compartment t

5.

Feedwater lines outside Visual Monthly When containment to tre reactor Operating feedwater pump chack valves A visual inspection for indications of leakage from all design basis piping break locations.

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l Amendment No. 7, 93 137 l

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LlMITING CONDIT10NS FOR OPERATlON SURVE1LLANCE REQUlREMENTS 3.6.I Shock Suppressors (Snubbers) 4.6.I Shock Suppressors (Snubbers) 1.

During all modes of operation The following surveillance except Cold Shutdown and Refuel, requirements apply to all safety all safety-related snubbers related hydraulic and mechanical listed in PNPS Procedures shall snubbers listed in PNPS Procedures.

be operable except as noted in 3.6.1.2 through 3.6.I.3 below.

The required visual inspection interval varies inversely with An Inoperable Snubber is a the observed cumulative number properly fabricated, installed of inoperable snubbers found and sized snubber which cannot during an inspection.

pass its functional test.

Inspections performed before that interval has elapsed may be Upon determination that a snubber used as a new reference point to is either improperly fabricated, determine the next inspection, installed or sized, the However, the results of such corrective action will be as early inspections performed specified for an inoperable before the original time snubber in Section 3.6.I.2.

Interval has elapsed may not be used to lengthen the required 2.

From and after the time that a interval.

snubber is determined to be inoperable, replace or repair Number of snubbers found the snubber during the next 72 inoperable during inspection or hours, and initiate an engineering during inspection interval:

evaluation to determine if the components supported by the Subsequent snubber (s) were adversely affected Inoperable Visual Inspec-by the inoperability of the Snubbers tion Interval snubbers and to ensure that the supported component remains 0

18 Months + 25%

capable of meeting its intended 1

12 Months ! 25%

function in the specific safety 2

6Monthsi25%

system involved.

3,4 124 Days t 25%

5,6,7 62 Days t 25%

Further corrective action for 8 or more 31 Days

~+ 257.

this snubber, and all generically susceptible snubbers, shall be The required inspection interval determined by an engineering shall not be lengthened more evaluation.

than one step at a time.

3.

From and after the time a snubber Snubbers may be categorized in is determined to be inoperable, two groups, " accessible" or improperly fabricated, improperly

" inaccessible" based on their installed or improperly sized, if accessibility for inspection the requirements of Section(s) during reactor operation.

These 3.6.I.1 and 3.6.I.2 cannot be met, two groups may be inspected then the affected safety system, independently according to the or affected portions of that above schedule.

system, shall be declared inoperable, and the limiting 1.

Visual Inspection Acceptance condition for that system Criteria entered, as appropriate.

A.

Visual inspections shall verify:

Amendment No. 20, 60,93 137a

LIMITfNG CONDZTZONS FOR OPERATTON SURVEILLANCE REQUfREMENTS 3.6.I Shock Suppressors (Snubbers) 4.6.I Shock Suppressors (Snuobers) 4.

Snubbers may be added to. or 1.

That there are no visible removed from, per 10 CFR 50.59, indications of damage or safety related systems without impaired operability.

prior NRC approval.

The addition or deletion of snubbers shall be 2.

Attachments to the reported to the NRC in accordance foundation or support with 10 CFR 50.59.

structure are such that the functional capability of the snubber is not suspect.

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Snubbers which appear INOPERABLE as a result of l

visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval provided that:

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1.

The cause of the rejection is clearly established and remedied for that particular snubber, and 2.

The affected snubber is functionally tested, when necessary, in the as found J

condition and determined OPERABLE per specifications 4.6.I.2.B., 4.6.I.2.C., as j

applicable.

3.

For any snubber determined j

inoperable per 1

specification 4.6.1.2, clearly establish the cause j

of rejection and remedy the j

problem for that snubber, and any generically i

i susceptible snubber.

J 2.

Functional Tests (Hydraulle and j

Mechanical Snubbers) j A.

Schedule t

At least once per operating t

cycle (18 months), a representative sample (10%

of the total of each typJ:

hydraulic, mechanical) of 3

snubbers in use in the plant shall be functionally tested, either in place or j

in a bench test.

For each j

snubber that does not meet j

the functional test acceptance criteria of j

Amendment No. 20, 60,93 137b

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LIMfTING CONDITIONS FOR OPERATION SURVEILLANCE REQUfREMENTS l

4.6.I Shock Suppressors (Snubbers) f of each snubber, the date at which the designated service life commences and the installation and maintenance i

records on which the designated service life is based shall be maintained.

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B.

At least once per cycle, the installation and maintenance records for each safety related snubber listed in PNPS Procedures shall be reviewed to verify that.the Indicated service life has not been i

exceeded or will not be i

exceeded prior to the next scheduled snubber service life review.

If the indicated service life will be exceeded l

prior to the next scheduled I

snubber service life review, the snubber service life shall i

be reevaluated, or the snubber shall be replaced or i

reconditioned so as to extend its service life beyond the date of the next scheduled

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service life review.

This reevaluation, replacement or reconditioning shall be indicated in the records.

C.

This Snubber Service Life Monitoring Program shall become

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effective July 1, 1982.

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l nandment No. 60,93 117d

(PAGES 137e THROUGH 1371 AREDELETED)

(TABLE 4.6.2 PREVIOUSLY ON PAGE 138A WAS MOVED TO PAGE 137) l Amendment No. 42, 60. 62. 93

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Bases:

i 3.6.G and 4.6.G Structu:al Integrity The Pilgrim Nuclear Power Station Inservice Inspection Program conforms to the requirements of 10 CFR 50 Section 50.55a(g).

Where practical, the inspection of ASME Section XI Class 1, 2, and 3 components conforms to the edition and i

addenda of Section XI of the ASME Boiler and Pressure Vessel Code required by 10 CFR 50, Section 50.55a(g). When implementation of an ASME Code required inspection has been determined to be impractical for PNPS, a request for relief from the inspection requirement is submitted to the NRC in accordance with 10 CFR 50 Section 50.55a(g)(5)(ill).

Requests for relief from the ASME Code inspection requirements will be submitted to the NRC prior to the beginning of each 10 year inspection interval for which the inspection requirement is known to be impractical.

Requests for relief from inspection requirements which are identified to be impractical during the course of the inspection interval will be reported to the NRC on an annual basis throughout the inspection interval.

Certain ASME Code Class 1, Category B-J pressure retaining welds-have been designated as Group I welds.

These Group I welds shall be included in the sample of Class I welds requiring inspection during each ten year interval.

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Amendment No,19, 93 149

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Bases:

3.6.H and 4.6.H High Energy Piping Outside of Containment Analyses performed and submitted to the AEC as Pilgrim Nuclear Power Station, Unit #1, FSAR Amendment #34 Indicate that certain modifications to the station increase the protection against the potential effects'of postulated high energy piping failures outside the primary containment.

In order to provide greater assurance that the integrity of the high energy piping outside-the primary containment is maintained at an acceptable. level'in the interim until these modifications can be completed, an increase in the frequency of inspections of the areas of concern has been initiated.

The monthly visual inspection of high energy piping outside the containment while the station is operating provides greater assurance of the timely detection of postulated piping failures and allows appropriate corrective action to be performed.

Reference to Subarticle IWA-5240 of the ASME Boiler and Pressure Vessel Code, I

Section XI, 1980 Edition, Winter 1980 Addenda, ensures that appropriate visual examination techniques are used to implement the requirements of Technical Specification Table 4.6.2.

These visual examinations will normally be made with the indicated piping and insulation in its operating condition.

Subsequent to the completion of the modifications, the inservice inspection requirements defined in Section 4.6.G of these Technical Specifications will provide adequate inspections to allow timely detection of~ postulated failures.

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f Amendment No. 7,93 150

IAS S:

3.6.I & 4.6.I SHOCX SUPpRESSCRS (SNUBEERS)

Snubbers are designed to prevent unrestrained pipe motion under dynamic loaf s as might occur during an earthquake or severe transient, while allowing normal ther:al 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 seistic or other event initiating dynamic loads.

It is therefore. required that all snubbers required to protect the primary coolant system and all other safety related syste=s or components be operable during reactor operation.

The visual inspection frequency is based on maintaining a constant level of snubber pectection to systems. The cumulative number of inoperable snubbers detected during any inspection interval is the basis for establishment of the subsequent inspection interval and the existing inspection interval should remain in effect until its completion.

When the cause of the rejection of a snubber is clearly established and remedied for that snubber and verified by inservice functional testing, that snubber may be exe=pted fro: being counted as inoperable.

Generically susceptible snubbers are these which are of a specific make or =edel and have the sa=e design features directly related to rejection of the snubber by visual inspection, and are exposed to the same environ = ental conditions such as tamperature, radiatica, and vibration.

When a snubber is found inoperable, an engineering evaluation is initiated, in addition to 'the determination cf the snubber mode of failure, in order to deter =ine if any safety-related cocponent or system has been adversely af f ected by the inoperability of the snubber.

Initiating this evaluation within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ensures that prompt corrective action will be afforded.

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

The service life of a snubber is evaluated via manufacturer input and information through consideration of the snubber service conditions and associated installation and =aintenance records (newly installed snubber, seal replaced, spring replaced, in high radiation area, in high temperature area, etc...). The requirement to monitor the snubber service life is included to ensure that the snubbers period-ically undergo a performance evaluation in view of their age and operating condi-tio ns. These records will provide statistical bases.for future consideration of snubber service life. The requirements for the maintanance of records and the snubber service life review are not intended to af fect plant operation. Due to the number and complexity of the relevant interacting factors necessary to develop a comprehensive Service Life Program, this program shall become effective July 1, 1982.

151 Amendment No. ??, 60, 93

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3.

Special Reports Special reports shall be submitted as indicated in Table 6.9.1.

6.10 RECORD RETENTION A.

The following records shall be retained for at least five years:

1.

Records of facility operation covering time interval at each power level.

2.

Records of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.

A 3.

Reportable Event Reports.

4.

Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

5.

Records of reactor tests and experiments.

6.

Records of changes made to Operating Procedures.

7.

Records of radioactive shipments.

8.

Records of sealed source leak tests and results.

9.

Records of annual physical inventory of all source material of-record.

B.

The following records shall be retained for the duration of the Operating License:

1.

Record and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.

2.

Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

3.

Records of facility radiation and contamination surveys.

4.

Records of radiation exposure for all individuals entering radiation control areas.

9 5.

Records of the service lives of all hydraulle and mechanical snubbers listed in PNPS procedures including the date at which j

the service life commences and associated installation and maintenance records.

Amendment No. )@, 68. 88, 93 224 y

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