Text

Proposed Tech Specs Modifying App a of License DPR-35 Re Average Power Range Monitor Setpoints,Duration of Integrated Primary Containment Leak Rate Test & Auto Blowdown Timer
	ML19241B737
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Site:	Pilgrim
Issue date:	07/23/1979
From:	; BOSTON EDISON CO.
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Attachment A Technical Specification Change to APRM Setpoints Proposed Change Reference is made to Pilgrim Station Operating License No. DPR-35 Appendix A.

The bases for Specifications 2.1.A.1 and 2.1.B contain information pertaining to the APRM High Flux Scram (Run Mode) trip setting and the APRM Rod Block (Run Mods) trip setting as follows:

2.1.A.1 Bases (Page 16)

"The scram trip settir.g must be adjusted to ensure that the LHGR tran-sient peak is not increased for any combination of MTPF and reactor core thermal power.

The scram setting is adjusted in accordance with the formula in Specification 2.1.A.1, when the maximum total peaking factor is greater than 3.06 (8X8 array)."

2.1.B Bases (Pages 17 5 18)

"As with the APRM ecram trip setting, the APRM rod block trip setting is adjusted downward if the maximum total peaking exceeds (3.06) for 8 X 8 fuel, thus preserving the APRM rod block safety margin."

The desired changes consist of:

In 2.1.A.1 bases delete "3.06 (8X8 array)" and add the following:

"....the design value.

This adjustment may be accomplished by increasing the APRM gain and thus reducing the slope and intercept point of the flow referenced APRM High Flux Scram Curve by the reciprocal of the APRM gain adjust."

In 2.1.B bases delete "3.06" and replace with "the design value," and add "As with the scram setting, this may be accomplished by adjusting the APRM gain."

Reason for Change Pilgrim Nuclear Power Station Technical Specifications require the adjustment of the APRM high flux scram and APRM rod-block settings whenever the actual maximum total peaking factor in the reactor core exceeds a specified value.

The requirement to adjust these settings ensures that adequate safety margins are maintained for all core power distributions.

The current Technical Specifications do not address the method to be used to accomplish the secting adjustment of the APRM circuits.

This amendment would clarify the Technical Specification requirement by changing the bases to identify the method to be used to adjust the APRM circuit settings.

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a Attachment A (Cont.)

Safety Considerations Changing the gains in the APRM amplifiers is equivalent to adjusting the set points because in either case, the circuits would perform their safety function in an idertical manner for identical conditions in the reactor core.

These changes do not present any hazard considerations not described or implied in the license application as amended.

These changes have been reviewed by the Nuclear Safety Review and Audit Committee and reviewed and approved by the Operations Review Committee.

Clarification Changing the gains of the APRM amplifiers effectively changes the scram set-points.

The technical justification of this method is presented below to demonstrate that the Technical Specification requirement is satisfied exactly.

To this end, a realistic set of conditions will be used as an example:

Core Power 20%

Recirculation Flow 40%

Core TPF 5

As per Specification 2.1.A.1, the scram setpoint is:

S=

( 0. 65 b' + 5 5 )

FF

-. MTP F --

For 8XS fuel S =

(0.65)(40) + 55 3.06

= 49.572 5

Therefore, the core power must increase a facter of 49.572/20 = 2.4786 to cause a scram.

(Note that the power increase cannot be caused by a flow change or affect peaking factor to stay within the constraints of the Technical Specification.) This scram requirement is accomplished by changing the APRM gain so that the APRM indicates a factor of TPF/3.06 greater than actual core power.

For this example, af ter the APRM's have been adjusted, they will have a scale factor of 0.612.

APRM = (Actual Core Power)z TPF 3.06

= (20%)

(5) 32.68

=

(3.06)

Scram occurs when APRM indicates:

APRM =.65 (40) +55 = 81%

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Attachrent A (Cont.)

Therefore, power must increase a factor of 81/32.68 = 2.4786 to cause a scram exactly satisfying Technical Specification requirements.

The actual value of the peaking f actors for 8X8 fuel was re=oved and replaced with "the design value" to save future Technical Specification changes of the bases as these values change.

Schedule of Chance This change will be put into effect upon receipt of approval by the commission.

Boston Edison Ccapany proposes that pursuant to 10CFR170 this is a Class III amendment.

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Attachment B Technical Specification Change concerning duration of Integrated primarv Containment Leak Rate Test (ILRT)

Proposed Change Reference is made to Pilgrim Station Operating License No. DPR-35 Appendix specification 4.7.A.2a p.

153.

The desired change would consist of replacing the existing paragraph:

The test duration shall not be less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for integrated leak rate measurements, but shall be extended to a sufficient period of time to verify, by measuring the quantity of air re-quired to return to the starting point (or other methods of equivalent sensitivity), the validity and accuracy of the leak rate results.

With the following:

The test duration shall~not be less than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for inte_ rated leak rate measurements, but shall be extended to a sufficient period of time to verify, by measuring the quantity of air re-quired to return to the starting point (or other methods of equivalent sensitivity), the validity and accuracy of the leak rate results.

Reason for Change The current specification requires 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of outage critical path time be extended for leak rate testing, due to the fact that the original wording of this specification was made at a time when leak rate measurements and relrulations were accomplished through manual methods, and a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months test was considered mandatory for accurate results.

However, at present, enhancements in computer data aquisition and data reduction have since invalidated the preceeding mandate, and a test duration of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ia a more realistic approach for compliance to the technical specification.

Safety Considerations Reducing the minimum duration of the IPCLRT from 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months does not present any hazards to the public health and safety.

This change has been reviewed and approved by the Operations Review Committee and reviewed by the Nuclear Safety Review and Audit Committee.

Clarification 1.

The functional integrity of the primary containment can be demon-strated independent of the test period, as long as equilibrium has been obtained in the measured variables prior to commencement of the test.

This is consistent with 10CFR50 Appendix J.

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a Attachrent 3 (Cont.)

2.

A recalculation of the 1976 IPCLRT at Pilgrim has been conducted and statistical leak rate has met acceptance criteria on the basis of a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months test.

In fact, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months calculation results are more con-servative than the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months test.

Calculated results of mass point leakage rates including a 95% upper confidence limit of.561%/ day versus.412%/ day for the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months test and the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months test, res-pectively, meet the acceptance criterion of.715%/ day.

3.

Data acquisition, reduction and calculational capability have been enhanced since the original IPCLRT was conducted in 1972.

Data collection and alaysis for the original IPCLRT in 1972 was at a frequency of 3 data sets per hour, present capability as used in the 1976 test is 6 data sets per hour.

The increased yield of data coupled with off-line computer analyses has given enhanced validity to statistical test results.

4 NRC regulations for IPCLRT, 10CFR50, Appendix J, specifies that statistical leak rate be adjusted to that which would be obtained for a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months test, but does not require an actual 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months test period.

5.

ANSI N45.4-1972 specifies that a test period of shorter than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months may be used if "It can be demonstrated to the satisfaction of those responsible for the acceptance of the containment that a leak rate can be accurately determined during a shorter test period.

Schedule of Change This change will be put into effect upon receipt of approval by the Commission.

Boston Edison Company proposes that pursuant to 10CFR Part 170 this is a Class III amendment.

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ATTACHMENT C Technical Specification Change RE:

T. S.

3.6.1 Introduction Pursuant to 10 CFR 50.59 Boston Edison hereby propose,.he following changes to Appendix A of License No. DPR-35.

Pro,osed Change Table 3. 6.1 (page 137c) is a list of all safety related snubbers.

This change consists of (a) deleting snubbers SS-2-10-17, SS-2-10-lS, SS-3-3-1 and (b) changing the designation of snubber SS-6-10 to S.

n-10-10 and snubber SS-2-20-5 to SS-2-30-5, (c) changing the elevation c:

SS-2-20-1, SS-2-20-2, SS-2-20-3, SS-2-20-4 to 42' and (d) changing the p r e fix SS to S for snubbers located outside the drywell. Attachment A is the proposed Table 3.6.1 that reflects these changes.

Reason for Change Snubbers SS-2-10-17 and SS-2-10-18 and SS-3-3-1 are not in the plant.

SS-2-10-17 and SS-2-10-18 wer e removed when the reactor recirculation pump discharge valve 4" bypass lines were removed.

This change was reported in the 1976 annual 10 CFR 50.59 report.

SS-3-3-1 was removed when the control rod drive system return line was removed.

This change was reported in the 1978 annual 10 CFR 50.59 report Thus these changes update the Technical Specification to account for previously documented plant modification.

The remaining changes are changes in snubber nomenclature and are made for administrative clarity.

Safety Considerations These Technical Specification changes relfect plant modifications whose safety considerations have been previously documented.

Hence the health and safety of the public is adequately protected when these proposed Technical Specifications become effective.

This change has been reviewed and approved by the Operations Review Committee and the Nuclear Safety and Review Co=mittee.

Schedule for Change This change will be put into effect upon receipt of approval from the Commission.

Fee Determination Boston Edison Company proposes that pursuant to 10 CFR 170 this is a Class II Amendment since it reflects changes that are adninsitrative in nature.

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ATTACllM F.Nl' A Table 3.6.1 SAFETY RELATED S110CK SUPPRESSORS (SNUlillERS)

Snubber No.

location Elevation Snubber in liigh Snubbers Snubbers Snubbers Radiation Area Especially Inaccessible Accessible During Shutdown Difficult to During Normal During Normal Remove Operation Operation I

SS-1-10-1 N in Steam Line 42' X (Drywell)

SS-1-10-2 Main Steam Line 42' X (Drywet.1)

SS-1-10-3 Main Steam Line 42' X (Drywell)

S S-1 4 Main Steam Line 42' X (Drywell)

SS-1-10-5 m in Steam Line 42' X (Drywell)

SS-1-10-6 Main Steam Line 42' X (Drywell)

SS-1-10-7 Main Steam Line 42' X (Drywell)

SS-1-10-8 N in Steam Line 42' X (Drywell)

SS-1-10-9 Main Steam Line 42' X (Drywell)

SS-1-10-10 W in Steam Line 42' X (Drywell)

SS-1-10-11 Main Steam Line 42' X (Drywell)

SS-1-10-12 Main Steam Line 42' V

(Drywell)

SS-6-10-6 Feedwater Sys.

41' X (Drywell)

SS-6-10-7 Feedwater Sys.

41' X (Drywell) 20 SS-6-10-8 Feedwater Sys.

44' X (Drywell)

SS-6-10-9 I'eedwater Sys.

41' X (Drywell)

SS-6-10-10 Feedwater Sys.

44' X (Drywell)

SS-10-30-1 RilR System 52' X (Drywell)

SS-10-20-2 RilR System 52' X (Dr se?

SS-10-20-3 RilR System 52' X (Drywe SS-10-20-4 RilR System 52' X (Dry i)

SS-10-30-5 Rilk System 24' X (Dr) ell)

SS-10-30-6 RIIR System 24' X (Drywell)

SS-10-20-7 RllR System 24' X (Drywell)

SS-10-20-8 RliR System 24' X (Drywell)

SS-10-3-9 RilR System 87' X (Drywell) s SS-10-3-10 RilR System 90' X (Drywell) g w SS-2-id-l Recir. System 42' X

X (Drywell)

SS-2-20-2 Recir. System 42' X

X (Drywell)

SS-2-20-3 Recir. System 42' X

X (Drywell)

' ) SS-2-20-4 Recir. System 42' X

X (Drywell)

- 'SS-2-30-5 Recir. System 15' X

X (Drywell)

" SS-2-30-6 Recir. System 15' X

X (Drywell) 5S-2-30-7 Recir. System 15' X

X (Drywel;)

C X

X

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ATTACil!!ENT A (cont.)

Table 3.6.1 SAFETY REl.ATED S110CK SUPPRESSORS (SNUB 15ERS)

Snubber No.

l.ocation El evat ion Snubber in liigh Snubbers Snubbers Snubbers Radiation Area Especially Inaccessible Accessible During Shutdown Difficult to During Normal During Normal Remove Operation Operation SS-2-30-8 Recir. System 15' X

X (Drywell)

SS-2-30-9 Recir. System 11' X

X (Drywell)

SS-2-30-10 Recir. System 11' X

X (Drywell)

SS-2-30-Il Recir. System 27' X

X (Drywell)

SS-2-30-12 Recir. System 27' X

X (Drywell)

SS-2-30-13 Recir. System 27' X

X (Drywell)

SS-2-30 '4 Recir. System 27' X

X (Drywell)

SS-2-30-15 Recir. System 27' X

X (Drywell)

SS-2-30-16 Recir. System 27' X

X (Drywell)

SS-2-20-19 Recir. System 16' X

X (Drywell) 20 SS-2-20-20 Recir. System 16' X

X (Drywell)

SS-2-20-21 Recir. System 19' X

X (Drywell, SS-2-20-22 Recir. System 16' X

X (Dryi' ell)

SS-2-50-23 Recir. System 17' X

X (Drywell)

SS-2-20-24 Recir. System 18' X

X (Drywell)

SS-2-20-25 Recir. System 16' X

X (Drywell)

SS-2-50-26 Recir. System 16' X

X (Drywell)

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ATTACllMENT A (cout. )

Table 3.6.1 SAFETY RELATED S110CK SUPPRESSORS (SNUBBERS)

Snubber No.

location Elevation Snubber in liigh Snubbers Snubbers Snubbers Radiation Area Especially Inaccessible Accessible During Shutdown Difficult to During Normal During Normal Remove Operation Operation SS-6-10-1 Feedwater System 42' X (Drywell)

SS-6-10-2 Femiwater System 42' X (Drywell)

SS-6-10-3 Fe ster System 42' X (Drywell)

SS-6-10-4 Fe t..

. iter System 42' X (Drywell)

SS-6-10-5 Feeawater System 42' X (Drywell)

SS-13-3-1 RCIC 38' X (Drywell)

SS-13-3-2 RCIC 38' X (Drywell)

SS-14-3-1 Core Spray 65' X (Drywell)

SS-14-3-2 Core Spray 65' X (Drywell) 20 SS-14-3-3 Core Spray 65' X (Drywell)

SS-14-3-4 Core Spray 65' X (Drywell)

SS-23-10-1 II. P. C. l.

42' X (Drywell)

SS-23-10-2 li. P. C. I.

42' X (Drywell)

S-23-3-30 II. P. C. I.

-3'09" X ll.P.C.I. Quadrant S-23-3-31 it. P. C. I.

-3'09" X ll.P.C.I. Quadrant S-23-10-32 II. P. C. I.

-3'09" X !!.P.C.I. Quadrant S-23-3-33

11. P. C. I.

-3'09" X 11. P. C.1.

Qua d r a n t N S-23-10-34 li. P. C. I.

-6' X ll.P.C.I. Quadrant CO S-23-10-35 H.P.C.I.

-6' X ll.P.C.I. Quadrant M S-23-3-36 li. P. C. I.

-3'09" X !!.P.C.I. Quadrant S-23-3-37

11. P. C.1.

-3'09" X 11.P.C. I. Quadrant S-10-3-43 RilR

-3'06" X RHR Pump Room pg

_. S-10-20-44 RilR

-3'06" X IUIR Pump Room S-30-3-45 RBCLN 83'5" X Reactor Building S-10-10-46 7:llR 6"

X Torus Compartment C

Tl Kidifications to this Table due to changes in high radiation areas should be submitted to the NRC as part of t he next 1icense amendment.

ATTACHMENT D ADMINISTRATIVE TECHNICAL SPECIFICATION CHANCE Proposed Change Reference is made to Pilgrim Nuclear Power Station - L' nit #1 Technical Spec-ification Appendix A, Section 6, Figure 6.2.2, titled "Pilgrin #1 Station Organisation".

The changes would consist of restructuring the chart (Exhibit A) to accommodate organizational changes at the Station.

Also, Section 6.5.A.2 "0RC Cc= position" will be updated to reflect another ORC member.

Reason for Change Recently, Boston Edison Company has initiated changes to improve the Health Physics Program at Pilgrim and as a result new positions have been created.

The subject positions cre titled as follows:

Chief Radiol <ical Engineer Senior ALARA Engineer ALARA Engineer ALARA Health Physics Technicians Senior Waste Management Engineer Prior to the creation of the Chief Radiological Engineer the Health Physics Program was under the responsibility of the Chief Technical Engineer.

It is Bos ton Edison's intention, by means of these new positions, to increase Station management's attention toward the Health Physics Program.

Safety Considerations Since the proposed changes affect only the Administrative Control Section of the Technical Specifications there are no safety-related changes involved.

These changes have been reviewed by the Nuclear Safety Review and Audit Connittee and have been reviewed and approved by the Operations Review Co=mittee.

Schedule of Change The described changes to the organi;ation are presently in effect.

Boston Edison Company proposes that pursuar t to 10CFR Part 170 this is a Class II Amendment.

Attachments:

Exhibit A Exhib i t B q

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EX11IBIT 3 s'

6.5 REVIEW AND AUDIT A.

OPERATIONS REVIEW COMMITTEE (ORC) 1.

FUNCTION The ORC shall function to advise the Pilgrim Station Manager on all matters related to nuclear saf ety.

2.

COMPOSITION The ORC shall be composed of the:

Chair =an:

Station Manager Member:

Methods, Compliance & Training Group Leader Member:

Chief Opere _ag Engineer Member:

Chief Technical Engineer Member:

Chief Maintenance Engineer Member:

Chief Radiological Engineer 3.

ALTER?'ATES Alternate members shall be appointed in writing by the ORC Chairman to serve on a temporary basis; however, no more than two alternates shall participate in an ORC quorum at any one time.

4.

MEETINC FREOUENCY The ORC shall meet at least once per calendar month and as convened by the ORC Chairman.

5.

OUORUM A quorum of the ORC shall cons st of the Chairman and two cembers including alternates.

6.

RESPONSIBILITIES The ORC shall be responsible for:

a.

Review of 1) all procedures required by Specification 6.8 and che.nges thereto, 2) any oth<r proposed procedures or changes thereto that affect nuclear safety.

b.

Review of all proposed tests and experiments that affect nuclear safety.

c.

Review of all proposed changes to the Technical Specifications.

d.

Review of all proposed changes or codifications to plant systems or equipment that affect nuclear safety.

e.

Investigation of all violations of the Technical Specifications and shall prepare and forward a report covering evaluation and recom-mendations to prevent recurrence to the Nuclear Operations Manager and to the NSRAC Chairman.

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ATTACHMENT E Prcro_ed Technical " ecification Change to Table 3.2.B, Auto Blowdown Timer Proposed Chance Reference is made to Pilgrim Station Operating Licenne No. DPR-35, Appendix A, Table 3.2.B, page 49.

The desired change would consist of replacing the Trip Level Setting Values of the Auto Blowdown Timer from the current 120 seconds t 5 sec. with the following values:

2 90,

$120 seconds.

Reason for Change A recent review of ECCS ar.alysis input parameters has shown that the upper limit of the ADS blowdown timer (125 sec) permitted by the present Technical Specifications is higher than the valae used in past and present ECCS analyses (120 sec).

This disagreement results in a 3 to 50 nonconservatism in peak clad temperature in the LOCA analysis.

Even though this temperature differ-ence is well within the margins existing at PNPS, it would be desirable to have the Technical Specification and LOCA input entirely compatible.

A change of input to the LOCA analysis would require a complete reanalysis which would not only be expensive but would unnecessarily reduce operating margins.

The proposed Technical Specification change is therefore the preferable action to assure compatibility between the Technical Specification and LOCA analysis.

Safety Considerations Reducing the response time of the ADS will increase the effectiveness of the ECCS by allowing LPCI to be operable earlier in the LOCA.

The timer limit of 90 sec. is long enough to allow HPCI's to start (designed for 25 sec.) or to allow the operator to cancel the ADS signal if the main control room in-formation indicates the signal is false or is not needed.

Since the input to the LOCA analysis will remain unchanged, this Technical Specification change will not affect the accident analysis.

Schedule of Change This change will be put into effect upon receipt of approval from the Co= mission.

Boston Edison Company proposes that pursuant to 10 CFR Part 170 this is a Class III amendment.

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ATTACHMENT F Technical Specification Change Concerning In Sequence Criticals Proposed Chance Reference is made to Pilgrim Station Operating License No. DPR-35 Appendix A, Specification 3.3.A.1 and corresponding bases p. 87.

The desired change would consist of adding conditions to the bases (attached) to allow shutdown margin demonstration by a method other than that currently specified in our Technical Specifications.

Reason For Change The two rod method (pulling the strongest worth rod and a diagonally adjacent rod to a specified position) produces a highly peaked flux distribution which maximizes the worch of the second rod.

This high worth can lead to sudden unexpected criticals with fast periods.

Conversely, the dispersed uniform withdrawal sequence is designed specifically to minimize the rod worths; thus the probability of a high reactivity incertion incident will be substantially reduced.

Safety Considerations This Technical Specification change will nc ' compromise the health and safety of the public since it of fers a more conser s'ive approach to demonstrating the shutdown margin, yet still retains the use o. the two rod demonstration if such is required in the future.

This change has beta reviewed and approved by the Operations Review Committee and reviewed by the Nuclear Safety Review and Audit Committee.

Schedule of Change This change will be put into effect upon receipt of approval by the Commission.

Boston Edison Company proposes that pursuant to 10 CFR Part 170 this is a Class III Amendment.

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3*.3 and 4.3 BASES:

c.

Reactivity Limitation The core reactivity limitstion is a restriction 4.

to be applied principally to the design of new fuel which =ay be loaded in the core or into a particular refueling pattern.

Satisfaction of the limitation can only be demonstrated at the time of loading and must be such that it will apply to the entire subsequent fuel cycle.

The generalized form is that the reactivity of the core loading will be limited so the core can be made suberitical by at least R + 0.25% ak at the time of the test, with the strongest control rod fully withdrawn and all others fully inserted.

The value of R in %4k is the amount by which the core reactivity, at any tiue in the operating cycle, is calculated to be greater than at the time of the check; i.e., the initial loading.

R must be a positive quantity or zero.

A core which contains temporary control or other burn-able neutron absorbers may have a reactivity characteristic which increases with core life-t ime, goes through a maximum and then d2 ceases thereafter.

The value of R is the difference between the cal-culated core reactivity at the beginning of the operating cycle and the calculated value of core R

reactivity any time later in the cycle where it

$;R would be greater than at the beginning.

The value

@ 0, of R shall include the potential shutdown =argin loss 6C assu=ing full B C settling in all inverted poison 4

tubes present in the core.

A new value of R must m,

be determined for each full cycle.

EE The 0.25%Ak in the expression R + 0.25%4k is provided as a finite, demonstrable, subcriticality margin.

This margin is demonstrated by full with-drawal of the strongest rod and partial withdrawal of an adjacent rod to a position calculated to in-sert at least R + 0.25%Ak in reactivity, or by an insequence, xenon free cold critical measurement to demonstrate at leas t R + 0.25% Ak in reactivity with the most reactive control rod fully withdrawn.

Observation of suberiticality in this condition assures subcriticality with not only the strongest rod fully withdrawn but at least an R + 0.2:% Ak margin beyond this.

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ATTACHMENT G Technical Specification Change To "IRM Channel Calibration" Ref. A) USNRC Region I letter to Boston Edison Company dated September 21, 1978.

Acknowledging response to Inspection 50-293/78-18 Proposed Change Reference is made to Pilgrim Station Operating License No. DPR-35 Appendix A, Table 4.1.2.

The changes would consist of adding further calibration and testing frequencies to the IRM High Flux Instrument Channel Section (see attachment).

Reason for Change This change as requested by the Commission through an I&E Inspec. ion and sub-sequent phone call (Ref. A) will serve to clearly specify the functional and calibration test requirements of the Nuclear Instrumentation Intermediate Range Monitoring (IRM) System.

Safety Considerations This change does not present any hazard considerations not described or im-plied in the license application as amended.

This change has been reviewed by the Nuclear Safety Review and Audit Committee and reviewed and approved by the Operations Review Committee.

Sc_hedule of Change This change will be put into effect upon receipt of approval by the Co= mission.

Boston Edison proposes that this request is exempt f rom any fee determirations since this amendment request was initiated by the Commission as clarif. cation to the affected Technical Specifications.

(b

TABLE 4.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENT CALIBRATION MINIMUM CALIllRATION FREQUENCIES FOR REACTOR PROTECTION INSTRUMENT CilANNELS Instrument Chai:nel Group (1)

Calibration Test (5)

Minimum Frequency (2)

IRM liigh Flux C

Comparisoa to APlet on Controlled Note (4)

Shutdowns rol1 Calibration once/ ope ra t iny, cyc 1 e APRM liigh Flux Output Signal B

lleat Balance Once every 3 Days Flow Bias Signal B

Internal Power and Flow Test Each Refueling Outage LPRM Signal B

TIP System Traverse Every 1000 Effcctive Full Power liours Iligh Reactor Pressure A

Standard Pressure Source Every 3 Months liigh Drywell Pressure A

Standard Pressure Source Every 3 Montha Reactor Low Water Level A

Pressure Standard Every 3 Months liigh Water Level in Scram Discharge Volume A

Note (6)

Note (6) ha>

Turbine Condenser Low V,cuum A

Standard Vacuum Source Every 3 Months O

E Main Steam Line Isolation Valve Closure A

Note (6)

Main Steam Line liigh P.adiation B

Standard Current Source (3)

Every 3 Months Turbine First Stage Pressure termissive A

Standard Pressure Source Every 6 Months Turbine Control Valve Fast Closure A

Standard Pressure Source Every 3 Months

^ Turbine Stop Valve Closure A

Note (6)

~)

Reactor Pressure Permissive A

Standard Pressure Source Every 6 Months i' )

u$

ATTAC WEN T H Technical Specification Change To Remove the Power Restriction Requirement For Testing MSIV Closure Time

Reference:

a)

Bcston Edison Coapany letter to NRC dated March 22, 1978 " Proposed Tech-nical Specification Change to the Main Steam Line High Flow Setpoint."

b)

NRC letter to Boston Edison Company dated September 19, 1978, " Amendment No. 34 to Facility Operating License No. DPR-35 for the Pilgrim Nuclear Power Station, Unit No.

1" Proposed Change Reference is made to Pilgrim Nuclear Power Station Unit #1 Operating License No. DPR-35, Appendix A, Specification 4.7.D.l.b(2) p.

160.

The desired change would consist of removing the current power restriction requirement of 50% teactor power when verifying closure time of Main Steam Isolation Valves.

Reason For Change In our letter, Reference (a) we proposed a change to the Technical Specification for Pilgrim Nuclear Power Station, Unit No.

1, related to the trip level setting for the High Flow Main Steam Line instruments.

The purpose of that modification was to alluw testing of the MSIV's and Turbine Stop Control Valves at higher (full) power levels, thus eliminating the n~eed to reduce power for testing pur-poses.

This request was granted by the Commission per Reference (b).

However, our Ictter Reference (a), failed to include all the Technical Specifications that would be affected by the subsequent amendment, and as such now submit this proposed change.

Safety Considerations This proposed change was intended to be included in Amendment No. 34 and there-fore the Safety Evaluations that accompanied Reference (a) and (b) satisfy all concerns related to safety.

This change has been reviewed and approved by the Operations Review Committee and reviewed by the Nuclear Safety Review and Audit Commi t t e e.

Schedule of Chaage This change will be put into effect upon receipt of approval of the Cocnission.

Fee Determination Boston Edison Company proposes that pursuant to 10 CFR 170 this is a Class II Amendment.

I 7md-

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ML19241B737

Text

Reference:

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