Text

Refueling & Spent Fuel Storage Information & Monthly Operating Report for May 1978
	ML19095A366
Person / Time
Site:	Surry
Issue date:	06/13/1978
From:	Stallings C; Virginia Electric & Power Co (VEPCO)
To:	Hartfield R; Office of Nuclear Reactor Regulation
References
	Download: ML19095A366 (52)
	v • d • e

,

VIRGINIA ELECTRIC AND POWER COMPANY :-~i RICHMOND,VIRGINIA 23261 June 13,

1973:

Mr. _R. _A. Hartfield, Acting Director Office of :Management Information and Program Control U. S. Nuclear Regulatory Commission Washington, D.. C.

20555

Dear Nr. Hartfield:

Serial No. 33'9.

PO&M/DLB:das Docket Nqs~ 50-280 50-281.

License Nos. DPR-32.

DPR-37

?~. :!.;

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--ic:

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Enclosed are the following reports for Surry Power Station Unit Nqs. 1 and

2.

Attachment,1:.

Attaclnnent 2: - :

Refueling and Spent Fuel Storage Information, Surry Unit i>!o. 1 Refueling and Sp~nt Fuel Storage Information, Surry Unit. l'i!o. 2 Monthly Operating Report, Surry Power Station, Unit Nos.* 1 and 2;. for :May 1978:.

Very truly.yours,

~.fi'l.xff"au~v-~

C. M. Stallings Vice President ~ Power Supply and Production Operations Enclosures (3 copies) -

cc:

Dr. Ernst Volgenau, Director (10. copies)

Office of I~spection and Enforcement 1'.1r. James P. O'Reilly, Director (1 copy)

Off ice -of Inspection and Enforcement j

('I

..f, ATTACHMENT 1

1.

Name:. SURRY UNIT NO. 1

2.

Next scheduled date for refueling shutdown:

11/1/79

3.

Scheduled date for restart following refueling: 5/1/80 (refueling outage extended for steam generator replacement).

4.

Technical Specifications Changes:

No changes are forseen due to the reload core design. (The review by both the station and system safety committees will take place prior to submittal of the reload.license to NRC.)

S.

Scheduled date forreload licensing submittal:

2/1/80

6.

Important licensing considerations:

None

7.

Number of fuel assemblies (as of 6/1/78):

a.

In the core:

157

. b.

In the spent fuel storage pool:* 436

8.

Licensed spent fuel pool storage capacity:*

a.

Current:

1044

9. Projected date of last refueling that can be discharged to the spent fuel storage pool assuming currently licensed capacity:

October, 1984.

Surry Units No. 1 and 2 share a coIIDD.on spent fuel storage pool.

r-

ATTACHMENT 2 I

1.

Name :

SURRY UNIT NO. 2

2.

Next scheduled date for refueling shutdown:

10/8/78

3.

Scheduled date for restart following refueling: 4/15/79 (refueling outage extended for steam generator replacement)

4.

Technical Specifications Changes:

No changes are forseen due to the reload core design (The review by both the station and syst'em safety committees will take place prior to submittal of the reload license to the NRC).

5.

Scheduled date for reload licensing submittal:

1/1/79

6.

Important licensing considerations:

None

7.

Number of fuel assemblies (as of 6/1/78)

a.

In the core:

157

b.

In the spent fuel storage pool:* 436

8.

Licensed spent fuel pool storage capacity:*

a.

Current:

1044

9.

P~ojected date of last refueling that can be discharged to the spent fuel storage pool assuming currently licensed capacity: April, 1985.

Surry Units No. 1 and 2 share a common spent. fuel storage pool.

e ATTACHMENT 3 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION MONTHLY OPERATING REPORT REPORT NO.

78-05 MAY, 1978 Superintendent - Station Openkions APPROVED:

Y" MANAGER

-i-CONTENTS Section Operating Data Report - Unit fll 1

Operating Data Report - Unit ff 2 2

Unit Shutdowns and Power Reductions - Unit Ill 3

Unit Shutdowns and Power Reductions - Unit ff 2 4

Load Reductions Due to Environmental Restrictions - Unit ffl 5

Load Reductions Due to Environmental Restr.:i:ctions - Unit ff 2 6

Average Daily Unit Power Level - Unit Ill 7

Average Daily Unit Power Level - Unit /12 8

Summary of Operating Experience 9

Amendments to Facility License or Technical Specifications 11 Facility Changes Requiring NRC Approval 12 Facility Changes That Did Not Require NRC Approval 12 Tests and Experiments Requiring NRC Approval 15 Tests and Experiment~ That Did Not Require NRC Approval 15 Other Changes, Test and Experiments 16 Chemistry Report 17 Description of All Instances Where Thermal Discharge Limits 18 Were Exceeded Fuel Handling 19 Procedure Revisions That Changed the Operating Mode Described in 22 the FSAR Description of Periodic Tests Which ~ere Not Completed Within 23 The Time Limits Specified in Technical Specificatibns Inservice Inspection 24 Reportable Occurrences Pertaining to Any Outage or Power 26 Reductions Monthly Operating Supplmement Sheet - Mechanical Maint. Unit #1 27 Maintenance of Safety Related Systems During Outage or Reduced 28 Power Periods Unit Ill

/

-ii-Section Monthly Operating Supplement Sheet - Mechanical Maint. Unit# 2 Maintenance of Safety Related Systems During Outage or Reduced Power Periods - Unit #2 Monthly Operating Supplement Sheet - Electrical Dept. Unit #1 Maintenance of Safety Related Systems During Outage or Reduced Power Periods - Unit #1 Monthly Operating Supplement Sheet - Electrical Maint. Unit #2 Maintenance of Safety Related Systems During Outage or Reduced Power Periods - Unit #2 Monthly Operating Supplement Sheet Instrument Maint.- Unit #1 Maintenance of Safety Related Systems Bu~ing Outage or Reduced Power Period - Unit 1 Monthly Operating Supplement Sheet - Instrument }faint. Unit #2 Health Physics Report Of Radioactive Effluents Sensitivity Data Procedure Deviations Reviewed by Station Nuclear Safety and and Operating Committee After Time Limit Specified in T.S.

Page 31 32 33 34 36 37 38 39 40 41 42 45 46

l.

~

"".'1-OPERATING DATA REPORT OP.ER.A.TING STATUS

. : 1. UnitName: *--------~S~ur~ry,,,....~l'°'"~--~~~~~,..,,------

0001 1soso1 2400 180531

2. Reporting Period:

3. Uc:nsed Thermal Power (MWt):

4. *Nameplate Rating (Gross MWe):

5. Design Elee:rical Riiting (Net MWe):

6. Maximum Dependable Capacity (Gross MWe)~

7. Maximum Dependable Capacity (Nee MWe):

2441 847.5.

822 811 775 Notes.

DOCKET NO.

DATE COMPLETED BY*

T:EUPHONE 50-280 6... 1-78 0.J. Costello 804-357-3184

8. If Changes Oc:~r in Capacity Ratings (Items Number 3 Through 7) Since Last Report. Give Reasons:

NA

9. P?wer L:vel To Wbic:h Restricted;lf Any (Net MWe):

10. Reasons For Restrictions, If Any:

NA' N/A

.,_

This Month

11. Hours In Reporting Period

12. Numbe*r Of Hour.i Reactor Was Critic:il

.. 13. Reactor Reserve Shutdown Hows

. : -14. Hours Generator On-Line

15. Unit Reserve Shucdown Homs

.. 16. Gross Tnennal Energy Gene:ated (MWH) * *

17. Gross Electrical' Energy Gene::ited (MWH) l~. Net Electric:tl.Enera,.y Gene::ited (MWH).

19. Unit Service F:ictor

20. Unit A..-ailability Factor

21. Unit Capacity Factor (Using MDC Net) ll. Unit Cap:ic:ity F:ictor (Using DER Net)

23. Unit Forc:ti Out:ige ~te

(.*

0 0

O 0

0 0

Yr.-t~Date 3,623

. 2, 667 0

2666.5 0

6,486,752 2,160,910 2,057,327 73.6%

73.6%

73.3%

69.1%.

0

24. Shutdowns Schedui~ Over Next 6 Months (Type. D:ite, and Dur:ition of E:u:h ):

25. If Shut Dovs."11 At End Of Report Period. Estim:ned O:ite of Startup:

26. Units In Test Status (Prior to Commercml Oper:aion):

INITIAL CRITlCALlTY INITIAL ELECTRICITY.

COMMERCIAL OPERA TtON 6/30/78 Forecist'

~ :

.Cumulative 31983.9 0

.71,,627. 748 2_3.520, 753 65 5%

. *65.5%

60.4%

57%

"16.8%

.. Ac:hi~ed

-~

r

.. OPERATING DATA REPORT

~TLNGSTATUS

. ; 1. Unit Name: -------~S~u~r?ry~2~~~~~~~~~~~--

0001 780501 2400 780531

2. Reporting Period:

3. Licensed Thermal Power (MWt):

4. *Nameplate Rating (Gross MWe):

5. Design Elecmctl ~ting (Net MWe):

6. M~um Dependable Capacity (Gross MWe):

7. Maximum Dependable Capacity (Net MWe):

2441 847.5 822 811 775 Notes DOCKET NO.

DATE COMPL.ETED sy*

TEUPHONE 50-281 6-1-78 O.J. Costello 804-357-3184 S. If Changes Oc:ur in Capacity Ratings (Items Number 3 Through 7) Since Last Report. Give Reasons:

N/A

9. P?wer L.evel To Which Rcscicted;lf Any (Net MWe):

N/A N/A

10. Reasons For Restrictions, If Anv:

.

This Month

11. Hours In Reporting Period

12. Number Of Hours Rc!:lctor Was crltictl

13. Re:ictor Reser.-e Shutdown Hours

. : -14. Hours G.merator On-Llne

15. Unit Reserve Shutdown Hours

** 16. Gross Thermal Energy Generated (MWH) *

17. Gross Electrical' Energy Gene::ited (MWH) l~. Net Electtic::tl.E.'lergy Gene::ited (M'WH).

19. Unit Semc:: F:ictor

20. Unit A,*ailability Factor 744 616.9 0

614.9 0

1,493,577 482,750 459,395 82.6%

-82.6%

21. Urut Capacity F:11:tor (Using MDC Net)

. 79. 7%

22. Unit C.ap:icity F:ic:tor (Using DER Net) 75.1%

23. Unit Forced Oucige R.lte 17.4%

Yr.~t~Date 3 623 2.993.8 2;988.2 0

7,267,966 2,371,315 2,253,597 82.5%

82.5%

80.3%

75.7%

4.1%

24. Shutdowns Schedu!~ Over Next 6 Months (Type. D:ite. :md Dur:ition of E:ich ):

Refueling and Steam Generator Inspection - Oct. 15. 1978. 5 weeks.

25. If Shut Down At End Of R~port Period. Estim:lted D:lte of Startup:

26. Units In Test St.:ltu)l (Prior to Commerci:ll Oper:nionJ:

INITIAl. CRITICALITY lNITlAL ELECTRICITY COMMERCIAL OPERA. TlON N/A Forecst"

~ ;

. 44,, 567 :.

29,406 6 28, 917 :i**

0 66., 978' 200.

21,938,709 64.9%

64.9%

60.2%

56.8%

23.8%

. Achieved

~...

Dale 78-1 780501 F: forced

S: Schtduled.

('1/77)

UNIT SHUTDOWNS AND POWER REJ)UcnONS REPORT MONUI May, 1978 rrL e:

0

=-

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0 ff ti Licensee Ev

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'II Reputl #

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s 744 c

1 Reason:

A*Equlpmenl failure (Explain)

B-Maintenance or Test C -Refueling

.: ~...

i

. D*Rcgulatury Restriction E-Operatur Training & License Examination F-Administrative G-Operatlonal Error (Explain) ll-01her (Explain) 3 Method:

I-Manual 2-Manual Scram.

3-Aulumallc Scram.

4-0ther (Explain) 50-280 DOCKET NO.

UNITNAME Surri DATE 6-1-78 COMPLETED DY G1 Kao~

TELEI HONE 804-3:,i7-31B!l Cause &. Currecllve

-Action ht Preveht Recurrence

. Unit shutdown for the entire month.

Refuell~& complet~d, turbine repairs

in progress, steam generator m~intenance in pro~ress.

4 s

EXhlbll G - lnstruclluns for Preparation of Dala.

Entry Sheets for Licensee Event Report (lERl File (NUREG*

0161)

Exhibh I

S:un~ Sourct

~.

......... ~

I w I

e.

f--'

N I

N 00 I.....

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Dale 78.-2 780524 F: forced

S: Scheduled (tJ/77)

UNIT SHUTDOWNS AND POWER REJ)UCTIONS

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J :c o-REPORT MONTII Licensee Event Reputt #

F 129.1 H

1 LER 78-019/03L

-0.

Reason:

A-Equipment Failure (Explain)

D-Maintenancc oi Test

(-Ref uellng

. D-Regulalory Restriction E-Operalor Training & License Examination F-AJminis1ra1i*1e G-Opcralional Error (Explain)

I I-Other (Explain) 3 May 1 1978 zz Suport Method:

I-Manual 2-Manual Scram.

J*Aulumallc Scram.

4-0lher (Explain)

A.-:

DOCKET NO.

50-281 UNIT NAME DATE COMPLETED llY TELEI HONE Surry 2 6-1-78 G. Kane 804-357-3184 Couse le. Corrective

Action lo Prcveni Redurcnce Snubber Inspection 4

s Ex.hlbil G *Instructions for Preparation of Data.

Entry Sheets for Licensee Event Report (LERl File CNUREG*

0161)

Exhibit I

Saml? S111m:t a ;co

w

41. c

za 1 I.c-

I

e.

I-'.

N I

N

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I

< LOAD REDUCTIONS DUE TO ENVIRONMENTAL RESTRICTIONS UNIT NO. 1 MONTH:

MAY, 1978 DATE TIME HOURS LOAD 2 MW REDUCTIONS :. MW MWH REASON (None during this repor ing peric d).

l MONTHLY TOTAL 0

.> LOAD REDUCTIONS DUE TO ENVIRONMENTAL RESTRICTIONS UNIT NO. 2 MONTH:

MAY, 1978 DATE Til1E HOURS LOADz MW REDUCTIONS z MW MWH REASON (None during this report ng perioc ).

MONTHLY TOTAL I

0

.. DOCKET NO.

50-280 UNIT __

r ___ _

DATE May, 1978 COMPLETED BY O. J. Costello AVERAGE DAILY UNIT POWER LEVEL MONTH _____

Ma_y_,_1_9_78 ________ _

AVERAGE DAILY POWER LEVEL AVERAGE DAILY POWER LEVEL DAY

{MWe-net)

DAY

{MWe-net) 1 0

17 0

2 0

18 0

3 0

19 0

4 0

20 0

5 0

21 0

6 0

22 0

7 0

23 0

8 0

24 0

9 0

25 0

10 0

26 0

11 0

27 0

12 0

28 0

13 0

29 14 30 15

o 31 0

16 0

DAILY UNIT POWER LEVEL FORM INSTRUCTIONS

on this form, list the average daily unir power level in MWe-net for each day in the reporting month.

Compute to the nearest whole megawatt.

These figures will be used to plot a graph for each reportmg month. Note that by using maximum dependable capacity for the*net electrical rating of the unit, there may be occasions when the daily average power level exceeds the l 00% line (or the restricted power level line). In such cases, the average daily unit power output sheet should be footnoted to explain the apparent anomaly.

DOCKET NO.

50-281 UNIT II DATE May, 1978 COMPLETED BY O.J. Costello AVERAGE DAILY UNIT POWER LEVEL MONTH ____

M_a_y __,_1_97_8 __ _

AVERAGE DAILY POWER LEVEL AVERAGE DAILY POWER LEVEL DAY (MWe-net)

DAY (MWe-net) 756.0 17 753.9 2

756.8 18 753.3 3

752.3 19 753.3 4

751. 6 20 753.5 5

748.8 21 753.8 6

750.0 22 752.3 7

zso J 23

-751. 5 8

749.8 24 452.5 9

745.6 25 10 753.1 26 0

11 752.3 27 0

12 752.0 28 0

13 758.1 29 0

14 757.1 30 624.7 15 757.9 31 748.0 16 753.1 DAILY UNIT POWER LEVEL FORM INSTRUCTIONS On this form, list the average daily unit power level in MWe-net for each day in the reporting month.

Compute to the nearest whole megawatt.

These figures will be used to plot a graph for each reporting month. Note that by using maximum dependable capacity for the* net electrical rating of the unit, there may be occasions when the daily average power level exceeds the l 00% line (or the restricted power level line). In such cases, the average daily unit power output sheet should be footnoted to explain the apparent anomaly.

\\*.

SUMMARY

OF OPERATING EXPERIENCE MAY, 1978 Listed below in chronological sequence by unit is a summary of operating experiences for this month which required load reductions or resulted in sig-nificant non-load related incidents.

UNIT 1 May 1 May 2 May 3 May 4 May 6 This report begins with the unit in a refueling outage with head bolt detensioning in progress.

Part length control rod unlatching operation completed at 1830.

Lifted reactor vessel head at 1030.

Full length control rod unlatching operation completed at 1500.

Commenced fuel movement at 2340.

Fuel movement stopped due to high airborne activity in the containment.

Fuel movement restarted at 1538.

May 12 -

Fuel movement completed at 1105.

Surveillance capsule W was removed for evaluation.

May 13 -

Full length control rod latching operation completed at 1000.

May 15 -

Lower reactor vessel head in place at 1055.

Part length control rod latching operation completed at 2200.

May 31 -

This report ends with the unit still in a maintenance outage UNIT 2 with the reactor vessel head bolts tensioned.

There is still primary maintenance underway and major turbine maintenance underway.

May 1 This report begins with the unit at 100% power.

May 5 At 0910 load was reduced to 94% to clear a delta-T runback alarm.

At 1100 the unit was returned to 100%.

May 9 At 1540 load was reduced to 96% t.o allow removal of both Lo Pressure Drain Pumps for cooling line maintenance.

At 1927 the unit was returned to 100% power.

May 24 -

At 1200 a rampdown was started for snubber inspection.

At 1645 the generator output breakers were opened, the turbine tripped, and the reactor manually shutdown.

A unit cooldown was commenced.

At 2330 the unit was <350°F/450 PSIG.

The plant temperature was to be maintained at approximately 340°F.

The snubber inspection was begun.

Hay 29 -

At 1430 the RCS was heated up to >350°F/450.PSIG as per normal plant recovery.

The snubber inspection & repair having been completed. 5 At 2350 the reactor was taken critical and at 2400 was at 2 x 10-ru:J.ps in the intermediate range of nuclear instrumentation.

SUMMARY

OF OPERATING EXPERIENCE

. (CONTINUED)

UNIT 2 May 30 -

At 0149 the turbine was synchronized to the system and a load increase begun.

At 0700 the unit reached 100% power.

May 31 -

This report ends with the unit at 100% power.

-ll~

AMENDHENTS TO FACILITY LICENSE OR TECHNICAL SPECIFICATIONS MAY, 1978 On May 10, 1978, the Nuclear Regulatory Commission issued Amendment Nos. 41 and 40 to the Operating License for the Surry Power Station Unit Nos.

1 and 2, respectively, which are designated as Technical Specification Change No. 53.

This change is the result of an NRC request to add surveillance re-quirements for the emergency diesel batteries.

Of significance are the follow-ing changes:

1.

Addition of limiting condition for operation of eme~gency diesel batteries.

2.

Addition of tests, test frequencies and acceptance criteria for emergency diesel batteries.

The Nuclear Regulatory Commission issued an order April 28, 1978, which modifies Surry Unit No. 2 Operating License.

The order results from the NRC Staff's review of an error in the Westinghouse Electric Corporation LOCA-ECCS analysis.

The error discovered was in the fuel rod heat balance equation involving the incorrect use of only half of the volumetric heat generation due to metal-water reaction in calculating the cladding temperature.

Of significance, the order has the following provisions:

1.

As soon as possible, the licensee shall submit a re-evaluation of ECCS cooling performance calculated in accordance with the Westinghouse Evaluation Model, approved by the NRC Staff and corrected for the errors described herein.

2. Until further authorization by the Commission, the Technical Specification limit for total nuclear peaking factor (Fq) for the Surry Power Station, Unit No. 2 shall be limited to 1.81 for a steam generator tube plugging level of 20.8% or less and 1.79 for a steam.generator tube plugging level of greater than 20.8% but less than 25%.

3.

Until further authorization by the Commission, the licensee shall conduct the operating surveillance program described in its letter of April 17, 1978 where APDM surveillance will be performed above 85% for Unit No. 2.

e e FACILITY CHANGES REQUIRING NRC APPROVAL There were none during this reporting period.

FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL MAY, 1978 The following facility changes were implemented during the month of May.

None of the changes constituted an unreviewed safety question per 10CFR50.59.

Design Change

1.

2.

DC-76-29 Flbw Shutoff of Low Pressure Heater Drain Pump Miniflow Description - This design change incorporates the addition of a solenoid valve in the recirculation line which will eliminate cavi-tation through the miniflow line when flow through the pump is greater than 400 gpm.

Summary of Safety Evaluation This system is not addressed in either the FSAR or Technical Specifications since it is not safety related and does not affect any safety related system.

Conclusion There are no safety implications for this design change which effect the health and safety of the public.

DC-76-36 Manipulator Crane Load Cell Readout Unit 1

Description - This modification is to mount an auxiliary readout for 1

the Manipulator Crane load cell to enable the crane operator to moni-tor the load and evaluation continuously.

Smmna.ry of Safety Evaluation The probability of occurrence or the consequence of an accident or malfunction of equipment important to safety and previously evalu-ated in the safety analysis report is not increased because no pro-tective circuits will be altered.

A possibility for an accident or malfunction of a different type than any evaluated in the safety analysis report is not created because the proposed Design Change does not alter the function of the original equipment.

The margin of safety as defined in the basis for any Technical Specification is not reduced because no protective functions have been altered.

Conclusion This Design Change does.not constitute an unreviewed safety question.or change the basis of any Technical Specification.

FACILITY CHANGES THAT DID NOT REQUIRE NRG APPROVAL (CONTINUED)

Design Change Unit

3.

4.

DC-77-02 Manipulator Crane Improvements Description - Modifications have been initiated to eliminate the possibility of dropping a fuel assembly.

A control circuit was inserted to prevent lifting a hung up assembly and the de-signed gripper guidepins were installed.

Summary of Safety Evaluation 1

The possibility of an occurrence or the consequence of an accident or malfunction of equipment important to safety and prev-iously evaluated in the Safety Analysis Report is not increased because the addition of the interlock circuit does not bypass any existing safety interlock circuits and the installation of the new guidepins will not alter their function or performance of the gripper.

A possibility for an accident or malfunction of a diff ererit type than any previously evaluated in the Safety Analysis Report is not* created because the failure of the new interlock system will either render the system to its original configuration or prevent the withdrawal of a fuel assembly.

This will not create a possibility for a different type of accident.

The new guide pins will function as the previous ones have and thus will not create a possibility for a different type of accident.

The margin of safety as defined in the basis for any Technical Specification is not reduced because the proposed Design Change will add two safety features which will increase the margin of safety as discussed in Section 3.10.

Conclusion This design change does not constitute an unreviewed safety question or change the basis of any Technical Specification.

DC-77-07 Manipulator Crane Gripper Modification Description - Modify the manipulator crane gripper to provide power assistance with the engage stroke.

Previous configuration included power stroke only while disengaging.

Summary of Safety Evaluation The probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the Safety Analysis Report is,not in-creased because this change does not modify any existing safety features or interlocks.

A possibility for an accident or mal-function of a different type than any evaluated previously, in the Safety Analysis Report is not created because the gripper and its interlocks will function as they were designed.

The margin of safety as defined in the basis for any Technical Speci-fication is not reduced because no margin of safety for the manip-ulator crane is addressed in Technical Specification.

1

4.

s.

DC-77-07 Conclusion.

FACILITY CHANGES THAT DID NOT REQUIRE NRC APPROVAL

.(CONTINUED)

Manipulator Crane Gripper Modification The design change neither constitutes an unreviewed safety question nor changes the basis for any Technical Specification.

Unit DC-77-18 Modification to Manipulator Crane Gripper Fingers Description -

Due to redesigned fuel assemblies from Westinghouse, the interface of the gripper fingers and fuel assemblies must be modified to provide proper latching.

This was accomplished by machining the fingers to pnroide more secure latching.

. Summary of Safety Evaluation The probability of occurrence or the consequences of an accident or malfunction of.equipment important to safety and previously evaluated in the safety analysis is decreased because, after the elimination of the point of interference, the reliability of the gripper will be assured, and, also the handling tools used in the fuel handling operation are conservatively designed initially, and operate on a fail-safe basis.

A possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report is not created because the removed metal will be on the thickest part of the finger and will not compromise the strength of the thinner part of the finger where no relief is required.

Also, the thinner part of the finger, where the surface is vertical, is where the maximum static movement stresses are located. The margin of safety as defined in the basis for any Technical Specification is not reduced because no margin of safety for the Manipulator Crane Gripper Fingers is addressed in the Technical Specifications.

Conclusion This design change does not constitute an unreviewed Safety Question per 10CFRS0.59 or change the basis for any Technical Speci-fication.

6.

DC-78-02 Chemistry Hot Lab Gas Bottle Storage 1,2 Description - The storage and piping configuration for bottled gasses used in the Hot Lab presented potential safety hazards.

which were eliminated with the relocation of the storage area outside the building adjacent to the other bottled gases (N2, Hz).

Summary of Safety Evaluation This design change does not modify safety related equipment, involve any postulation addressed in the FSAR, or reflect on any margin of safety defined in Technical Specification.

Conclusion This design change presents no implications which should effect the health and safety of the public.

e

'llll'rsT AND EXPERIMENTS REQUIRING NRG APPROVAL MAY, 1978 There were none during this reporting period.

ST-54 ST-56 TESTS AND EXPERIMENTS THAT DID NOT REQUIRE NRC APPROVAL MAY, 1978 Terminal Block Verification Unit This test was conducted May 11, 1978, to obtain' information 1

to formulate a response to IE Bulletin 78-02.

Results of the test will be reported in a specific response letter.

Limit Switch Verification This test was conducted May 11, 1978, to obtain information 1

to formulate a response to IE Bulletin 78-04.

Results of the test will be reported in a specific response letter.

I. ~

OT R CHANGES, TESTS AND EXPERIMENTS MAY, 1978 There were none during this reporting period.

/

PRIMARY COOLANT ANALYSIS MAXIMUM Gross Radioact., µCi/ml l.19E-2 Suspended Solids, ppm 2.0 Gross Tritium, µCi/ml 3.56E-2 Iodine..,;.131, µCi/ml 2.45E-3 I-131/I-133 Hydrogen, cc/kg o.o Lithiilm, ppm 0.35 Boron-10, ppm +

538.6 Oxygen-16, ppm 6.0 Chloride, ppm 0.07 pH @ 25°C 5.38 SURRY POWER STATION CHEMISTRY REPORT MAY 1978 T.. S. 6. 6.A.11 UNIT NO. 1 MINIMUM AVERAGE*

l.02E-3 4.18E-3 0.2 0.7 2.07£-2 2.40E-2

l.17E-4 5.68E-4 o.o o.o 0.00 0.15 445.7 499.0 0.6 1.5 0.05 0.05 4.56

. 4. 97

+ Boron-10 = Total Boron x 0.196 Phosphate 19 580 NON-RADIOACTIVE CHEMICAL RELEASES, POUNDS T.S. 4.13.A.8 Boron UNIT NO.

MAXIMUM MINIMUM 1.52E-l 2.38E-3 0.7 o.o 2.14E-l 8.09E-2 5.69E-4 l.37E-4

.2527

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~*..--.-------

702 Chlorine 0.0 Remarks:

Unit #1 Refueling shutdown during entire month.

Unit /12 Shutdown 5-24-78 to 5-30-78.

2 AVERAGE 8.67E-2 0.3

1. 60E-l 3.46E-4

.1142 36.2 1.57 111. 7 0.000 0.05 6.88

. DESCRIPTION OF ALL INSTANCES WHERE THERMAL DISCHARGE LIMITS WERE EXCEEDED MAY, 1978 On Uay 30, 1978, a thermal excursion was experienced during startup of Unit II with maintenance in progress on circulating water fish screens and condenser waterboxes.

The excursion was reported in accordance with Technical Specification 4.14.C.l.

May 30, 1978 Exceeded 3°F/Hr Rate of change There were no reported instances of significant adverse environmental impact

FUEL HANDLING MAY, 1978 During this month Unit 1 was refueled.

There were 64 new fuel assemblies inserted into Unit 1 and 64 discharged to the spent fuel pool.

No new fuel was received or spent fuel shipped.

DATE.

T.PPED/RECEIVEI?

FUEL HANDLING mur No. i MAY,

'NO OF

.£\\SSE~..BLIES. PER SHIPl*!ENT ANSI NO *

. INITIAL ENRICH!IENT NONE l;)URJ NG THIS PERIOD.

1.*

NEW OR SPENT FUEL SHIPPING CASK ACTIVITY LEVEL 0

.j

)

i

. i I

i i

I DATE.

NO OF

?PED/RECEI'!EI)

ASSEHBLIES PER FUEL HAl'mLING mIIT NO 2

-~;. 1978 e ANSI NO.

SHIPMENT INITIAL ENRICII!'IENT NONE Dl RING THIS PERIOD.

NEU OR SPENT FUEL.

SHIPPING CASK ACT I VI Tl LEVEL PROCEDURE REVISIONS THAT CHANGED THE OPERATING MODE DESCRIBED IN THE FSAR MAY, 1978 There were none during this reporting period.

DESCRIPTION OF PERIODIC TESTS WHICH WERE NOT COMPLETED WITHIN THE TIME LIMITS SPECIFIED IN TECHNICAL SPECIFICATIONS MAY, 1978 There were none during this reporting period.

INSERVICE INSPECTION MAY, 1978 An eddy current, gaging inspection was performed in Unit #1.A,B and C Steam Generators in accordance with VEPCO NDT Procedure 14.1.. The Reg.

Guide 1. 83 inspection revealed 2 tubes in "A" S/G having defects >40%.

"B" S/G had no tubes with >40% defect.

"C'i.S/.G had 6 tubes with >40% *defect.

Greater than 12% of the tubes were inspected in all three S/G' s in accordance with.-Reg.

Guide 1.83.

The number of tubes schedul~d to be plugged are as follows:

A B

c Gaging 18 31 40 Reg. Guide 2

6 Approved W Pree edures were used in the repair of the 10 leaking plugs in "A" Steam Generator.

Westinghouse performed inservice inspections on the Unit #1 applicable safety related items detailed in Technical Specifications, Section 4.2-1 through 4.2-35.

This inspection included a visual on the reactor vessel interior surfaces, ultrasonic testing of the reactor coolant pump flywheel, visual and liquid penetrant on sensitized piping.

The detailed inspection program along with results will be published by Westinghouse and transferred to the site for records purposes.

The rough data is available on site for review.

No reportable indications were found in the above mentioned inspections.

Ultrasonic inspection was conducted on 20 welds in the low points of the C.S. and R.S. Systems in accordance with VEPCO letter serial number 345/112476, dated January 11, 1977.

No reportable indications were found in this inspection.

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l e REPORTABLE OCCURRENCES PERTAINING TO ANY OUTAGE OR POWER REDUCTIONS MAY, 1978 Unit #2 Outage occurred on May 24-30, 1978 because of a require-ment for re-inspection of hydraulic seismic supports (snubbers).

The details of this inspection and results are contained in LER 78-09/03-L-0 (050-0231).

l.

MAY, 1978 MONTHLY OPERATING SUPPLEMENT SllEET MECHANICAL MAINTENANCE UNIT 1 llAINTENANCE ORDERS COMPLETED SAFETY RELATED ITEMS 114

2.

DESIGN CHANGES BEING WORKED DC-76-29, 76-36~ 77-02, 77-07, 77-18, 77-29, 78-01,

3.

76-30, 77-08, 78-10, "76-14, 77-19, 75-55 PERIODIC TESTS PERFORI-lliD PT-39 PT-12 PT-22.4A PT-22.4C PT-46 PT-16.4 6

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1.

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2.

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1.

MAINTENANCE ORDERS COMPLETED SAFgTY RELATED. !TENS 7

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DESIGN CHANGES BEING \\mRKED~~n_c_-_7_7_-_o_a~~~~~~~~~~~~~~~~~~~~~~~~~~-

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1.

MAINTENANCE ORDERS COMPLETED SAFETY RELATED ITEMS

3.

PERIODIC TEST PERFORMED 0

34 I

~

0 I

.J HEALTH PHYSICS MAY, 1978 There was no single release of radioactivity specifically associated with an outage that accounted for more than 10% of the allowable annual values in 10CFR20.

There were 3 individuals who received single radiation exposures specifi~ally associated with Unit #1 Steam Generator work which accounted for more than 10% of the allowable annual values in 10CFR20.101.

/

~'aclllty; Surry Power Station I.

l.lQUlll RELEASES

~~urrrrs*

l.

Gross Rad1oact1vJt~ (ll:y~

(a)

Total Ile leas~

Curies (h)

Avr:. Concentration Ile leased

µCi/ml (c~

Maximum Concentration Released J1Ci7ml

2.

Tdtium (a)

Total Rtoleased Curies

~b~

Avll* Concentration Released 11Gl7ml

1.

lllssolved Noble Cases (a}

Total Release Curies

~b ~

Avi;. Concentration Released 11Cl/111l Gross Alpha Radinactivi ty (a)

Total Released Curies (h)

AVIJ. concent1*ation lie leased 11Ci/ml

5.

Vol. of Li<lUid to Disch. Canal Liters

6.

Vol. of Dilution Water 1.iters

7.

Isotopes Relea~ed Ml'C uCl/ml Curies 1-Dl 3 l( 10-7 1-132 8 x 10-6 l-133 1 x 10'"

I-134 2 x 10-0

I-135 ___

1;-JCTo::-*

10-T C!l-1311 9 ll.

Cs-136 6 x 10-*

Cs-137 2 x 10,

Cr;-130 Co-57 4 x 10 *

---co-Sfi 9 x 10-a

---co-60 3 x.10-*

Mn-511

]

l(

10""~

Na-211 3 x 10-*

Cr-51 2 x 10-*

Fc-59 5 x 10-*

Nb-95 1 x 10-11 lla-J.a-140 2 x }(\\-s

---11-j;.:.aa REPQll'f OF RAllIOAC'l'IVE EFFLUENTS Du~KET; 50-280 and 50-281

.J ANllAllY FI:;RnllARY MARCii 6.J21i-l 5.14E-2

4. 70E-2 2.3~E-9

2. 41E-'t0.

2.17E-10 5 14E-9

3. 47E-lff

2. /lE-10 9.27E+l

!.J'.lt: :-1"1

l. llE+:.!

3.54E-7

l. lOE-7 5.12E-l

2. 38E-l

l. lllE+O 3.67E-l 9.0BE-10 5,55E-9.

l.69E-9 l.16E-5 3.70E-7 l.86E-6 4.431l-14

1. 741l-15 8.58E..:15 3.37U+7 7.17E+7 3.39E+7 2.62E+ll 2.l3E+ll 2.17E+ll

l. l12E-2 8.96E-3 l.02E-2

l. 71E-4

8. 96E-4 4.20E-2

3. 72E-4 6.06P.-4 4.J9E-2 5.lSE..;4 5.331\\-4 4.20E-2

l. 38E-'4

6. 70E-4 5.19E-2

l. 731,!-2

2. 22E-'2 l.76E-2

1. JSl!-3 9.27E-5 3.39Jl-2

4. 65E.,-'.4 3.ll2E-2

2. 22E-4 l.l.l6E-2

1. 54E-4 C'SbE-4

1. 64E-3 l.14E-l "T.'5'4E-2 l.03E-l l.03E-l

6. OCiE-2

8. lJE-2 6.911E-3 2.11JE-'.l 5.21E-3

l. 59E-2 2.511E-2 5.51.E-l

3. 60E-2 6.16Erl 3.32E-2 9.09E-4

1. 511!-4

1. 06E-4 3.51E-3

8. 311E-4 3.0lE-'.l

1. lllE-3 7.S3E-4 9.25E-5

1. 36E-3

2. lOE-3

-rr-TT';7-*

~

-~**

APllll.

1. 63E-l 6.86E-10 5.67E-JO 8.13E+l 3.43E-7 9.97E-l 4.21E-9 l.'l2E-5 8 lOF.-14

_l.filfil_

2,3Zf'.:+ll

2. OSE-2 5.49E-4 s,orn-J

. -2..16E-4 l.14E-2 2.14E-2 1,

1. 3JE-4 l.40E-l 9.19E-2 9.51E-'.l 4.70E-4 3.97E-2

3. 33E-3.

PAG~ l Y EAll, __

1~9~7....,8'----

HAY 8.38E-2 4.96E-10 4.92E-10

2. 71E+l l.61E-7 l.06E-l 6 25E-10 1.62E-5 9.60E-14 5 OlF.+7 1 69E+ll 4 52E-3 l 88E-4 8.471~-3 2 90E-2 7, 60E-~

3.00E-1 2.33E-l 2.69E-2 9.47E-5

1. 99E-l 7.51E-4 1.421>-2 2.48E-5

___. _ ___._*-----:~-**--"'!.................,......... ~.*

. *.*. 'A, i

I

r.:

. ;.i::,::

{,,!1,

.r::

f~f

\\'.::.

.jl ~

l_......

~--.-~,.*-1-*~--*~""" -***~':**r,.*...

- - "**\\ I 1.

.'~ '*l'fhr* rtlt'.'f"' :~***') ;' * ***' *. 1

- ' "* ~"*.

l'ACI!.l'l'Y;'

I.

- TOTAL--- -rE'rror'"

--=i

~~

I l.97E:..5 Ar-4l 3 x 10-*

3.BlE-5

9'.3l:E-6.

k.

i<

Sr-89 3 x 10-*

11-------*-------.-------*-~

3

~* ~

59~E='--4~--1t-------1 Sr-90 3 x 10-*

C-14 B x 10-*--------~-----1---.,-----t-----..,----,--i--.,,-,---.,.-,..,--~---:-----i-------i-------ii------ 1-----

8.

P c rcent of lO~C_.F_* l_l-'2_0 __________

1

~P_,.e,..r_,.c"'e"'n_,_t_

1 ___

5_;_1_9_E_-_2 __ i-_3_._7l_E-_2 __

-ll __

1~*-l-"O.,.t:.,..:.,.,_i __

11_"'1"";"""7;.2E;.' --1~_

1

__..z~._.Z~Z,..E~-.... 2 __ 1 _______

11 _______ 1 ____ _

ll.

ATllllOUNE UEWASES

_Curi ea

. 9.471>+1"

~,80!'+2 '.

l 95t:+2..

l.2LiE+3 9.61E+l Cur:! es

1. 23E-4

~E-=r-1: '.l6E-4 S.34E-3 5.37E-2

l.

Total Nob le Gaues

2.

Tot11l Halogens

~.

Total Particulate Gross Radio-Curles 1.9'.lE-5 l.li7E-6"...

1 :6 i E.:.s

7.lJE-5

i. 291!-4 Curjes li.46E+o 4.:nE+O i". VE+2

6. 79E+O 3.64E+O

__ 2.._c ti vlc_y_Jfu'_L ____________,

~-

Total 'l'dtium Cur1!l§

.. l,Ot,E.:.6*".',

7.iiiJg.'..j'

--~:-t;2t*::./***

"i."5-7Ii:.::6 ****..J.: llE-5 S.

'l'otal Particulate Gro!Hl Alpha Ra<lioa~tlvJ.9'._ ____________

1~~~~-t------~t--""~=;,-1-..---1t~~

......._...,....~11....,..~~=.,..""--'-l-'--'--'-="-'"'--=--1-------l-------1-----

Ci.

Hnxiurum Hobie Gos ttelease Rate*

.ucJ I sec 6.37E+2 9 ~ 40E*t-2 1

i.;;oe+'.l

__ l_ 044E+4 6.51E+2

-7.

11crccnt of Appl1cable Limit for

'l'cchnl ca 1 Specif l __.c_a_t_i_o_n_s ______ 1 _____

11 _______ 1,___,,---,----i------*-------*-------o-------*i------*-----

---~~-,..-;:-~~--:~::-':i-'

~-',~"-*e-~-:-s""e""s __________

g. ~=~~:::~ ; t'~~~=;

'~:,j~~~!

f

~ :;?~=~

~'.~~~=~

~:~~~=i (c)

!'articulates

. l'erccnt I Ll9E-3 2.30ETI --1..!l.l'"'E-_...

4--t1-"'-9"'-"--96~E"-*-"'4"--!l-;::.l;...6"'s°'E'°"*-"'3=--i1-------l-------t-----

~-----------------*

"-~.~~,---.,------r.-.-------i:

8.

Isotopes Released; Cnrlcs

(_a~)

l'a1*tlcnlates Cs-114

2. 01.i,-s

_Ll1E-L-' _LlOE-9 __ 1---*~----fl-_.;.2.;... "-27'-'E'-'--'6'---i-------H-------t------1 Cs-135 1---*----f------1--,-------1------1 C;:!.::_1_3_7 ____________,____

l.53E-8 1.l12E-6 2,Jl1E-6 l.37£-6 7.41E-6 c~.. s_*-~1=3.1-'l _____________, _____ 0~~~:~=*--~ 2. 17E-3 J.26E-J

~C-~:_2g 3.0SE-~ ---2:..§.QE=It_ _L_l.:.1&:.5 7.97E-5

4. 72_E-_5 __, _______ -------n------

Co-60

~~

l. 161'-s

__hlliE=-5._1 __

,~* *~B~7~"~-~6 __ 1 __ ~9~. o_s_e_, __ s __ **------*------*----

__________ tln.::.:i!*___

3,~llE-g_,__

5.!SE-9 1,

2.49E-6

__________ fe-59

~

1---*-~---*i---*----1------ 1-------11-----*

=-~-liic.!lli~~----* --~--.~-~--~-*=*-~=-~='"""*~~ r-- ~ 1rn-s

.L5]E=4..

.. '*.. ~.

__... __.:___ _______ ~------..:.:..----:----......

~..........,.........

1.

(_.,

2. -

(

I

\\

Qy.

.I

\\

SEUSITIVITI DATA Minimum Detectable Activity a *.

b.

c.

d.

e.

f.

g.

Gross Alpha; S.27E-9

µCi/ml Tritium; 4.0E-6 µCi/ml Strontium-89; less than 3.0E+l pCi per total filter composite.

Strontium-90; less than 2.0E+o pCi per total filter composite.

C02-14; less than 7.0E-10 µCi/ml.

Stronti1Jlll-89; less than 9.0E-9 µCi/ml.

Strontium-90; less than 8.0E-10 µCi/ml.

Multichannel Analyzer - GeLi

a.

Liquid and/or Radiogases lOOcc Ba-La-140 2.0E-7 I-131 5.lE-8 Xe-133 5.4E-8 Xe-135 2.9E-8 Cs-137 9.2E-8 Cs-134 7.2E-8 Co-60 l.SE-7 Co-58 9.8E-8 Cr-51 3.SE-7 Mn-54 1.0E-7 Fe.-59 2.4E-7 Ar-41 l.6E-7 I:-134 1.lE-7 I-132.

l.OE-7 Na-24

1. 7E-7 I-133 6.9E-8 I-135 4.SE-7 Kr-85 L36E-5 Kr-88 5.SE-8 Kr-87 5.3E-8

Kr-85m 2.3E-7 Xe-138 4.7E;..9 Xe-135m 5.8E-8 Xe-133m

1. 7E-7 Rb-88 8.SE-7

b.

Radio iodines I-131 2.3E-6 I-133 3.lE-6 I-132 4.7B-6 I-134 4.SE-6 I-135 2.0E-5

Jo.~, r

)

PROCEDURIEVIATIONS. R~~;WED BY STATIOR ILEAR SAFETY AND OPERATING COMMITTEE AFTER TIME LL.'1IT SPECIFIED IN TECHNICAL SPECIFICATIONS MAY, 1978 Number Unit Title Deviation PT-26.4 1

Radiation Monitoring Step 5.2.4-Unable to Equipment Test During lGl.

Fuel Handling Operations Step 5.2.5-Unable to 1G9.

Step 5.3.4-Unable to lGl.

Step 5.3.5-Unable to lG9.

Step 5.4.4-Unable to lGl.

Step 5.4.5-Unable to lG9.

verify verify verify verify verify verify This procedure was deviated 05-01-78 and reviewed by the Station Nuclear Safety and Operating Committee 05-18-78.

alarm alarm alarm alarm alarm alarm PT-24.1 1

Fire Protection Water Pump Steps 5.12 through 5.17 and 6.1.2, not performed.

This proceduEe was deviated 05-09-78 and reviewed by the Station Nuclear Safety and Operating Committee 05-24-78.

ML19095A366

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ML19095A366

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Dear Nr. Hartfield:

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