ML19338G294

Submits Assessment of Util Capabilities to Implement near-term TMI Action Plan Requirements & Proposed Schedule for Implementation.Duration of Outages to Accomodate Mods Estimated to Be 2 & 3 Wks for Units 2 & 3,respectively

ML19338G294
Person / Time
Site:	Peach Bottom
Issue date:	10/10/1980
From:	Daltroff S PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Eisenhut D Office of Nuclear Reactor Regulation
Shared Package
ML19338G284	List: ML19338G285 ML19338G294
References
RTR-NUREG-0660, RTR-NUREG-660 NUDOCS 8010290025
Download: ML19338G294 (18)

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at 9 'D PHILADELPHIA ELECTRIC COMPANY i ii 2301 M ARKET STREET P.O. 80)5 8699  %

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L, 7 October 10, 1930 le: Docket Nos.30-277 50-278 ,

Mr. Darrell C. Eisenhut, Acting Director Division of Operating aeactors U.S. "uclear 'Te;ulatory Co mtai s s io n Washint, ton, DC 20553 SU3 JECT: I etp le m e n t a t i o n of NaC Action ?lan 3.aquirementa

Reference:

(1) NURIC 0574, TNI-2 Lessons Learned Task Force Status Report and Short-Tera Recommendstions (2) Corresponlance dated May 7, 1930, D. C. Eis.nhut, NRC, to All Operating Reactor Llcensees (3) Correspon. lance datad Septenbar 5, 1930,

3. C. 7.isenhut to All Licensees of Operating Plants.

Dear Mr. 51senhut:

This letter presents an ansesacent of our capabilities to implament near term NRC Action Plan r e q u i r e m e n t's ( N Un f.G 0650),

and a p roposed sche dule for implomancation. Attachment A to this letter describes the status of Phi'.adelphia . lectric Conoany's efforts to 1 nt:lu. ten t these requircucit t c . Che requirementu veru o ri::ina lly identified in referencen 1 and 2. The september 5, 1000 la:ter fron D. C. Eisenhut (reference 3) provided new desi,1r. I criterit for many of the requireuents in references 1 and 2, and pronosei a new implementation schedule. Attachnent A pronoses an implettestation schedule that is covisistent with r.any of the " ". ':

proposci scheduled chanacs, as n e l '. as some a d d i t ion.41 ch a n ;e s in I the schadule, specifically for P e a s:5 P,o t t o: ,.couic

• i* o w e r Ctation j Unit 2 2nd Unit 3. The additional changes are :riaarily a result l

i r . .Darrell C. Eisenhut P t; e 2 Octohce-10, 1930 of the inability of some vendors to meet requested delivery schedules, and an effort to minimize the nuaber of plant transients and economic i= pact reculting fron plant outajes, and would aerait laplomentation in n nore effective, orderly fashion.

A suamtry of the proposed schedulcofor the Peach sottoo Units is presented in accachsent 3.

We propose the followin; outage schedule for iaplemantation of near term Action Plan sodifications.

1) January 1, 1931: A Unit 2 outage commencin; on this date to ic71ement those Action Plan modifications not limited by e r:u i p m e n t unavailability.

2) On or before March 15, 1931: A Unit 3 outate to accommodate refueling and tu complete most near term Action Plan modifications requiring an outaJe. Equipacnt unavailability may preclude full implacentation of the followin; itcas:

II.F.l(3) Rich- Range Containaent Radiation 'f o nit o ra ,

11.7.l(4) Containacat Pressure Monitors, nn1 II.F.l(5)

Containment Water Level :tonitcrs.

3) on or before January 1, 1962: A Unic 2 outage to complots modifications for uhich equipt.cnt procurement nroblems pracluded their implementation during the January 1900 outa e.

Duration of outages to accoamodate the near tera nodifications is estinated to be two and three weeks for Peach cottom Jnit 2 2nd Unit 3 respectively. Current estiaates of the renlaceaent encr;y charges to area custoners associatec witn the three week Unit 3 outage is $23 aillion. In view of the fact that ?cach lotton Unit 3 is scheduled for an extended refueling /aodification outano starting in early :f arch 1911, vc propose a ralaxation of the Peach 3occom Unit 3 isolementatioa date fcr Action Slan requirements so that the modifications can be acccomodated during the scheduled refueling outage. In addition, our fuel vendor has advised us that in order to acet the desian shutdown cargin criteria. in the next ?cach :ottoc Unit 3 fuel cycle, Unic 3 must attain a shutdown exposure of at least 7700 Mt0/ TON. Obtaining this exoosu re will be difficult if a three-week outage oust be taken nrior to the scheduled refuelin3 shutdown. Given a three week outage, operation of Peach 3ottaa U2it 3 beyond'the currently scheduled refuelina outaac date of early :: arch 1931 vill be necessary, causing further unavailability of the Unit during the sumoce of 1931, or alternatively the nunher of fresh reload as s erablic s aay have to be reh cod which will degrade the energy available fron ?cach 3otton Unit 3 during its next cycle.

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tir.'Darrell G. 21senhut PaSe 3 October 10, 1980 A re-schedulind of the *C ni t 3 near term Action Plan

. requirements until the olanned refueling outage would (1) enhance-the iaplementation of the Unit 2 sodifications by avoidin3 potencial difficulties associated with manpower availability and productivity which are inherent sich late Deccaber work, (2) increase Unit 3 availability during the summer of 1931, (3) permit the implementation of Ac-ion Plan requirements that would not be oossible earlier.due to engineering and procurement restraints, and'(4) would save area custoners approxinately $20 million (equivalent to more chan one aillion barrels of oil).

This relaxation of the Unit 3 inplementation date would involve only ten weeks and does not compronise olant safety.

The proposed schedule provides for implementation of the NRC requirements in a reasonable fashion. Should you have any questions regarding this satter, olease do not hesitate to Contact it s .

Very truly yours, UCB: bas Attachments bec: V. S. noyer J. S. Xeoper J . L' . ' G a lla g h e r S. J. Eradley M. J. Cooney R. H. Moore

'J . T. Ullrich C. C. Mistner A. H. Sellers U. M. Alden/V. C. 31 rely

ATTACH:!ENT A PEACH 30TTOM ATOMIC POWER STATION IM?LEMENTATION STATUS OF NEAR TERM ?"tC ACTIO!! PLAN ITEMS

1. Reauirement: Shift Technical Advisor (I.A.I.1)

The Shift Technical Advisor (STA) shall receive training in plant design and layout, including the capabilities of instrumentation and controls in the control room. They shall also have received specific training in the response and

. analysis of the plant for transients and accidents. This la /el of training shall have been attained by January 1, 1931. A description of this training, and the long term STA program, including qualification, selection critoria, and training plans shall be submitted by January 1, 1981.

2esconse Saclosed is a partial responso to this request, and a proposal to complete the comprehensive STA training program presently in progress and scheduled to run checugh February 18, 1931.

Six candidates were selected f rota our enginee ring staff for the STA position earlier this year, and have been attending a full time, twenty-two week training course since September 2, 1930.

The t ra ining progran, which includes simulator t r a i n i n g, , closely parallels the proposed INPO training standard for STA's, and is described in attachment'C. The training curriculum exceeds the requirements identified in NUREC CS73, Short Term Lessons Learned.

Ne pro >ose that the assignment of these personnel to replace the interin STA's be deferred until cocpletion of the training discussed a bo ve , expected to be Fabruary 13, !"ll. This would avoid cisruption of the training efforts, therefore maximizing benefits gained by the trainces from the p r o r, r a m . . Information regarding the long term training and qualification criteria will he submitted January 1, 1981 as requested.

2. Reruirement: Plant Shielding (II.3.2)

Perform a radiation and shielding design review of the spaces a rc und systems that may, as a result of an accident, contain hip.hly radioactive materials by January 1, 1980. Completo socifications, based on the shielding study, to assure auequate access to vital areas following an accident by January 1, 1981.

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Attachment A (Cont *d)

Nesnonse This requirement parallels the plant shieldin?; s tudy o f M*JROG 0578, item 2.1.6b. The results of the shielding study were presented in our submittal of January 31, 1930, S. L. Valtroff to

!! . R. Denton. As a result of this study we proposed for completion by January 1, 1981, the relocation of equipment and facilities. This involves the relocation of the spent fuel makeup controls to areas outsido the reactor building; and the establishment ot . 'sckup ra d io c h o.a i s t ry laboratory at a distanco froc. the plant.

l The NI:C Region 1 :n c o t i n g , held in Arlington, VA, on September 22, 1980, provided additional clarification of the source torn design criteria for the plant shielding study. A reassessment of the sh'elding study, based on this new clarification, indicates that post accident radiation conditions will not impact on reactor building accessibility and the avai "cy of the present raciochemistry laboratory. Therefore, we propose that i

1.,plementation of the modifications described above he deferred until such time that their need is clearly established.

3. Renuirement: Post Accident dampling Station (II.3.3)

Upgrade the capability to obesin samples from the reactor coolant system and containment atmosphere under high radioactivity conditions by January 1, 1901.

Ressonso To provide for equipment delivery, and installation in an orderly fashion, we believe the January 1, 1982 implementation dato proposed in the September 5, 1930 letter from D. G. Eisenhut to be appropriate for this requirenetc.

4 Recuirement: Safety-Rslief Valve Qualification Testina (II.D.1)

A plant specific submiccal for safety.and relief valves is required by July 1931.

lesnonse The Peach Soccom type safety and relief valves are included la che neope of the prototype q ua li f i ca t ions tunting t. o be perforued under_the auspices of the 3'.m Owners Group. 'le are providing the

Attachaent A (Cont d) necessary support through the Owners Group to develop and comploce the testing program. The best effort for the Owners Group to complete the qualification testing is July 1, 1931.

Additional time will be necessary to evaluate the daes and provide a plant specific submitta*. He propose that the schedule presented in correspondence dated September 17, 1980, D. 3.

Wacors, Chairman of the SWR Owners Group, to R. !! . Vollmer, NRC, be conaidorad as an acceptable schedule to satisfy this requirement. The proposed schedule is as follows:

Complete test facility: December 31, 1930. Complace shakedown tests: February 15, 198L. Complete operability tests: July 1, 1931. Complete test eupnets:

Decembe r 31, 1981.

5. Raouirement: Safety-2elief Valve Position Monitors (II.D.3)

Recctor coolant system relief and safety valves shall be provided with a positive indication in the control room derived from a reliable valve position detection device by J a r,u a r y 1, 1980. A qualified installation is required by J a r.u a r y 1, 1981.

Ressonse A reliaele direct positica indication syacem, utilising acoustic sensors, is presently op4 rational on all Peach Doctom safecy-relief valves. As stated la the November 21, 1979 letter f rom S. L. Daltroff to H. R. D e i. c o . . we sre in the process of upgrading this system to meet the safety grade design criteria applicable to this requirement. This task requires an outage on both units. We are prepared to isolement the improvements by January 1, 1981; however, for the reasons discussed in the cover leccce, we propose completion of all work during a Unit 2 oucale starting January 1, 1931, and a Unte 3 refueling outanc starting on or before :ta rch 15, 198L.

6. Tecuirement: Dedicated Hydrogsn Penetrations (II.E.4.1)

F. va lu a t e the design of the pur;o system for post accident combustible gas control of the containment atmosphere; and complete modifications, if reqaired, by January 1, l981.

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Attachment A (Cont'd)

Resnonse i The modifications to implement this requirement involve additional containment isolation valves on the Containment Acaospheric Dilution (CAD) system. This work has. been completed on Unic 2, while Unic 3 will require a scheduled outaae. For the reasons discussed in the cover letter, we propose implementation on Unit 3 during the refueling outage starting on or before :! arch 15, 1111. The June 30, 1931 Laplenentacion date proposed in the Septemmer 5, 1930 letter from D. C. Cisenhut is thereforo an appropriate schedule.

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7. Requirement: High Range Effluent tioni t o r (II . F .1(1) )

Provide high range effluenc monitors for noble gases by January 1, 1931 in accordance with the design criteria presented in the October 30, 1979 letter from H- 2. Denton re;;arding clarificacion of NUREC 0573, short Term Lessons

Learned.

Resoon9_e This raquirement parallels item 2.1.3b of NUREC 0573, Short Ters Lestana Learned. Three new monicering systems were installed earlier this yea.- to meet the NUaIC 0573 requirements. A letter from 2. W. Reid, NRC - Division of Licensing, to E. C. Sauer, states cSat Philadelphia Electric Company has satisfied the !!3C requiremects related to Item 2.1.3.b of the T:tI-2 Short Torm Lessons Learned requirements and Item III D.2.1 of the PfI action Plan (NUREC 0660). In light of the revised requirements for this system specified in section II.F.l(l) of the September 5, 1930

1. .om D. C. Eisenhut, it is not cicar at this time whether any further action regarding these monitors is required by Philadelphia C1ectric Company. We prosose that the modifications previously implemented, remain as an acceptabic res,onse to the requirement for upgrading the noblo gas monitors.' !!ow e ve r , if further modifications are required to meet the proposed criteria peesented in section II.F.l(l) of the September 5, 1930 letter, the time required to select, order, receive and install the systems would not oermit completion by the N2C proposed isolementation date of October 1, 1931. Therefore, we propose a deferral of the industry-wide implementacton date for section II.F.l(l) requirements until July 1932.

I Attachment A (Cont'd). I l

8. 9. c n u i r e m e n t : Sampling and Analysis of Plant Offluents (II.F.l(2))

Capability for effluent monitoring of radiolodines for the accident condition shall be provided with sa.npling conducted by absorption on charcosl or other media, followed by on-site laboratory analysis by January 1, 1931.

Ressonse l

3ased on the results of the shielting study submitted on January

31, 1930, S. L. Daltroff to H. R. Denton, we proposed the relocation of the iodine offluent sampling system from the reactor hvilding to ch'e turbine building by January 1, 1911, to meet the requirements of section II.F.l(2). As a result of additional clarification of the source term critoria providad at the N90 Region I meeting on Septenber 22, 19.10, and discussad in item 2 above, relocation of the iodine monitors is no longer deemed necessary co. meet this requirement. The Sentauber 5, 1910 lottar from E. C. Eisenhut presen

s dest:n critoria for the high ran3e radiciodine sampling syst.  : ,e representa -w t requir:3ents. Tha present i n c e .t l' a c io n at 'anch 'atton, which providss continuous effluent sampling for iodines and particulates, would require new equigment to comply with the now l criteria. The NRC proposed implenentation dato of October 1, 1981 wsuld not provide sufficient time because of the long lead times axpected for procurement of the naw custos equipment after an engineering review. Therefore, we propose an implementation date of July 1, 1932 to implement the new requirements for the radiciodine sampling system.

9. acquirement: Drywell Radiation :tonitors (II.F.l(3))

Install high range radiation monitors in the drywell by January 1, 10:11.

Rossonae f

Delivery of the monitors is p r e s e r. c ly scheduled for late this year (ano half of the monitors by November 15, 1930, and the other half by December 15, 1980). This modification could be completed during an outage s ta rt ir.c Janua ry 1, 1931, if the equimment is received on schedule and satisfactorily passes receipc inspection; however, a qualified recorder wil1~not be available by this date. Since cae schedule is very tight and a orderly installation is desirable, and qualified recorders will not be availably by January 1, l H.1, we believe that the October-D* lD lD N aM JJu\ hlb

t

~Actachmenc'A (Cont'd) i f 1, 1931 implementacion data propo sed in the September 5, 1900 lecter from D. G. Eisenhut will be necessary to effectively '

conolete implementation. Under this time schedule, implementation would be complaced on Unic 3 during the refuellag outage, and on Unit 2 before October 1, 198L.

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10. Retuirement: Containment Pressure Monitor (II.F.l(4))

Inicall high range containment pressure monitor by January l, 19J1.

Rosnon.:o Me have been actively working with the Geneen1 Electric Cnnpany and chu stut Owners Group for the past year to develop qualified eq.;pment to implement this reautremenc. There is no manufacturer of pressure transmiccers that we havo found that can meet the requirements of IEZi Standsed 323-1974.. We have contracted vich the conoral Cicceric Company to provide us with qualified equipment. However, due to problems with sub-vendor qualification programs, they are not able at this time to identify a delivery date for this equipment. Since chere is no

! existing containment pressure inscrumentation capable of monitoring the range required by the NRC, we hava purchased pressure transmitters from Rosemount that are qualifiad to ISEE Scandards 323-1971 end 344-1973. We are prepared..wich NRC

approval, to install chese Rosecount pressure cransmitters during the first scheduled outages proposed in the cover letter. This i

is the best available equipment on the narket today. Thorofore, we propose chat this 4odification be accenced as the permanent installation for upgrading the containment pressure instrumentation. Qualified recorders may not be available from che Ceneral Electric Company by the first scheduled outsgos. Ve propose to install non qualified racorders at this time and replace them vich qualiflad recorders during the first schedulad j outage following delivery. .

ll. Requirement:

Containment 4acce Level >tonitor (II.F.l(5))

Install high range containment level monitor by January l, 1991.

9 Tessonse 9e are experiencing the same procurement problems for this equipment as described for the pressure monitor in ices 10.

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Accachment A (Cont'd)-

Mowevor, the existing non-safety related containment water level ,

instrumentation is capable of monitoring the ranca required by the NRC. We propose upgrading the containment water icvol instrumentation during the first scheduled outage following delivery of the qualified equipment.

12. Raouironent: Containment Hydrogen Monitor (II.T.l(5))

j Continuous Lndication of hydrogen concentration in tho j containment atmosphere shall be provided in the control room.

tesnonso i

he original NRC design critoria for the hydrogen mo'icors, pecsonced in the October 30, 1979 clarification lettar on NUaEC 0573, required by January 1, 1931, a nessurement capability over

the range of 0-10% hydrogen concentration for both positive and negative ambient pressure conditions, and compliance uith j 3egulatory Guido 1.97, aevision 2. We have reviewed the d e s i,; n of the existing Containment Atmospheric Dilution (CAD) Systea analyzers installed at Peach Gottom, and conclude that they ,

comply with this criteria. On this basis,' modification to the

hydrogen analyzers is not required.

The September 5, 1930 letter provided several new requirements l for the hydrogen analyzers and proposed a revised coupletion date of October 1, 1931. The Peach Bottou equipment meets the now

]'

requironents except for the revised measurement accuracy requirement. . We propose that the accuracy requirement should bc deleted for the following reasons:

1_

l a) qualified, safety grade, hydrogen analyzers are not

! commercially available with an accuracy of 10.1 volume percent hydrogen for a 10 volume percent range.

b) The Peach ilottom containments t. r e inerted (maintained at less than 4% oxygen).

c) Post LOCA combustiule :;a s concentrations are con t roliert - by the' CAD System. The system is operated to add nitrogen and vent containment gases in order to maintain oxygen ccacentra't ion below the combustible liuit. Therefore, the CAD System oxygen analyzers are important for proper ccabustible gas control and the hydrogen analyzers are used fcr information only. (Refer to the Peach 3ottom P S A:t ,

S u p pletie n t .1, response to question 14.6 for further irformation.)

O e i

Attachment A (Cont'd) d) The post-accident samplind system being installed in resoonse to II.3.3 is designed to take containment gas samoles for gas chromatographic analys_J in t7e on-site laboratory.

The information regarding the hydrogen analyzers requested by October 1, 1931 in the September 5, 1930 letter will be provided by that date. Qualification of the oxygen analyzers is being pursued as part of the response to IE Bulletin 79-013.

13. Requirement: Auto Restart of RCIC (II.K.3.13)

The RCIC system InLtistLon logic should be moel i f l e il su that the TCIC system will restart on low water level by April 1, 1931.

Resoonse We are planning te implement this modification dur 13 the scheduled outane on Unic 2 to taplement Lessoon f.e a . ,e d requiraments, and on the Unit 3 refueling outage startina Starch 1, 198L, well before the implementation schedule of April 1, 1931.

14 Requirement: HPCI/RCIC 3reak Detaccion (II.K.3.15)

The pipe break detection circuitry should be modified so that pressure spikes resulting from !!? C I and RCIC system inttiation will not cause inadvertent system isolation.

Resoonse We are prepared to implement this modification by January 1, 1931. An outage is not required to implement the nodifications.

Ilowever, if implemented without an outage, it requires removing safety related equipment from service during installation. In additicn, plant availability may be jeopardi:ed by this work.

Therefere for this reason, and for reasons stated in the cover letter, we propose completion of this task during a Unic 2 outa;c starting January 1, 1931, and a Unic 3 refueling outs;e starting on or before March 15, 1931.

15. Recuirement: Technical support Center (III.A.L.2)

Upgrade the emergency su, port facilities in accordance with NURIC 0696 by April 1, 1982.

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. Attachment A (Cont'd)

Resnonse Item 2.2.2b, NUREG 0573, Short Term Lessons Learned, presented the requirements for establishing a permanent Technical Suoport Center _(TSC) by January 1, 1931. Correspondence dated January 2 1330, S. L..Daltroff t o !! . 2. Denton, oresented our-commitment _to meet this requirement by January 1, 1931.Section III.A.l.2 of the September 5, 1930 letter from E. G. Eisenhut, envokas NUREC 0696 as the dosign criteria for the TSC and proposes an implementation dato of April 1, 1932. We have submitted conments on NUR?.C 0696 (draft) in correspondence dated September 23, 1980, V. S. 3 oyer to S. L. Ramos, NRC. It is our understanding that NUREC 1696 will be issuud later this year.

i Additional time, as proposed in the September 5, 1930 letter, will be necessary to implement the new requirements anticipated in NUR2C 0696, and to complete in an orderly fashion our previous commitaents. 'J i t h the exception of data acquisition, we propose a completion date of April 1, 19dl, for the TSC. Philadelphia Electric Company's capability of implenenting the data

acqaisition-and other new.requiraments will be assessed following issuance of the final draft of NUROC 0696.

! 16. Requirement: Containment Isolation Dependability (II.E.4.2) l a) All non-essential systems shall be autouatically isolated by the custainment isolation signal by July 1, t

1931.

b) The containment setpoint pressure that initiates

! containment isolation for non-essential penetrations must be reduced to minimum, compatible with normal operating conditions by July 1, 1931.

Resnonce .

a) As stated in the September 5, 1930 letter, additional guidance will be provided by NR2 on the classification of essential vs..non-essential. Upcn receipt of r ' dance, we will intetate an engineering evaluation to identify moCifications, if necessary, to the containment isolation system. The time to complete the Engineering and procure new e q t. i p m e n t , and the need for a plan: outage, may preclude implementation by July 1, 1931. Na propose a deferral of the official impicsentation schedule until the NRC has reviewed the January 1, 1931 submittal from each licensee.

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1 Attachment A (Cont'd) b) In response to the Septenher 5, 1980 letter, we'have initiated studies of the feasibility of loweri.; the

' isolation setpoint and expect to report on this topic by January 1, 1981 as requested. If modifications are requirol as a result of the studies, the implomontation schedule of July 1, 1931.may not provide sufficient time to procure ocuipment and to accommodate the modifications during plant outages. We propose a' deferral of the official itaplomontation schedule until.the .IR C has reviewed the January 1, 19d1 suhaittal fron each ?lcensee.

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!* i'H PO S E D I f1P LIMil'.l; f AT i o ; S Cill !*til.E FOR I;E A R IT. H *l f.kC ACTION P l. A 's 1 T E M S Presuant P ropom?J ?!KC P ts i l a al etal.a L luc t r ic NRC I up l ease sit a t . on tar _ s. J S c tsa.lia l u

, lieplementatton S c ise dia l e Action P l .a n _ N n . Title t ; lie.l u l u ( J / 5 /ilu 1.c a t o r ) Unit ti ai t t i 5.A l.1 Stil f t Teclinf ca l Advisor 1 / 1 / 18 1 1/1/51 2/21/31 ?4 ./dl -

11.8.2 Plant S h t e l J i n r, 1/1/81 1/1/31 l lote I  !! o t e- 1 -

11.8.1 Pust Accident S a ss p l i n g itatton I/I/d1 I / I fil2 1/I/82 I/1/82

' I l . 8) .1 S.a t' e t y-P e l i e f Valve QualifIcatton Testing 7/1/81 1/1/81 1/4/62 I/1/82 1 1 . 11 . 3 S.i f e t y-P.e l l e t va l ve Punttton :Iun t t or 4/1/81 1/8/ul I/t/81 3/lS/u!

I l . it . 4 .1 noitcatei fly t ro,;e n renetrattuns 1/l/nl 6 / 30 / til couplete J/15/ul ll.F.l(l) ti l ;;te it a n g u Effluent !!on i t or 1/1/31 10/1/31 '4 u t e 2  ?;ote 2 II.r.l(2) ludine :Iunitor I/t/51 l oi l / tli 7 / 1 / 18 2 7/1/82 II.F.l(l) Contalunent R.adiation ttunttor 1/1/31 10/1/61 10/1/d1 1/13/31 II.F.l(4) Containr ent P r e s u u'r e Honitor 1/1/81 1/1/31 t/1/ul 3 /15 / til -

(:a u t e 3) ( *;o t e 1)

II.F.l(S) Containaent ifa t e r 1.e v e l tl.eni t or 1/1/81 1/1/31 ?to t e 4  :;ote 4 ll.F.l(6) Co n t a t s-. se n t tt yd r o;;u n 8:onitor 1/1/31 10/1/81 e n / t /..e en/s/ne II.K.1.13 Auto Restart of R C '. C 4/1/01 4/1/81 1/1/61 1/15/81

. II.K.l.ls lircl/HCle areak tutection 1/s/al I/1/u! I/1/ul 3/15/ul

P r e s e si t l' r o p o u c J lit:C P a l l a d u I pla l .a t. le c t r i c NHC Iuplessentarlun PropuscJ ScheJule l asp l e me n t a t l oin S c lie J u l e .

Actlun tIan No. Title Schedule (

__9/5/B0 B.e t t e r l tin i t 2 Uut t _3 III.A.I.2 Teclinics! Suppor& Center I/t/dl 4/1/82 '4 / t /8 8 4/1/38 thote.3) (Note ))

i Note la Huassusas.una leased on the NRC clarification of the plant shielding source terus indicates p r e s c en t .

p l .a ss t .ic s i g an is satinfactory.

Note 2: Clarification from HkC necessary. See itou 7 of Attaclemesat A.

Note 32 installation of non qualified recorder - Unit 2: 1/1/31, Unit 3: 3/15/81.

Qualified recurJer - i;.se scheduled outar.e folluuing delivery.

tiute 4 Non quallfled instruuentation presently installed. QualificJ i nns t ruue nt a t i on - first scheJuncJ oistago f.a l lois i ng d e l i v e r y .

1.nte 5: Except for data a cig u i s i t i o si aa n.1 o t tie r n ets r e q u i r cinc ei t s l ei .'10 :4 0 C 0 6') 6 .

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.ATTACHMC:lT C C U!!'I F. M T S il I I'T T C C!!N I C AL ADVIS0't TRAIMIMC P R O C 2 Ai!

The instruction includes five phases of training over 22 weeks. These are:

Phase I - Academic Training (o weeks)

Phase II -

Management / Administrative Controls Training (2 weeks)

Phase III - Plant Systems Training (3 weeks)

Phase IV - Accident Analysis Training (3 weeks)

Phase V -

Simulator Training (3 wooks)

Phases I -

IV are being presented at Peach 3ottom while Phase V, Simulator Training, will take place at the L i..tc r i c k Training Center.

Classroom. portions of the progras normally run 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per. day with about 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> por day allott1d for quizzes, examinations or structured study. There is at least one examination per class week. The details of each phase of the prograu are outlined as follows:

Phase I -

Sasic Academic Phase ( f3 weeks)

This portion of the program is a condensed version of the course normally prouanted to candidates for the reactor operator's licenaa. The overall objective is to provide the student with a basic understanding of the scientific and engineering principles of reactor plant operation. %cy academic fundamentals normally not included in a college curriculum are stressed.

Phase 11 - Manatoment/ Administrative Controls Phase 12 weeks)

This chase of the training introduces the duties and r e s p c '.s i bili t ie s of the Shift Technical Advisor. The objec tives are to provide prerequisite leadershis skills as wel,1 as an orientation on general plant operations and s fety to ensure that each STA is familiar with plant asnagement and administration. Phase II topics include the following:

Duties & Responsibilities of the STA Leadership Interpersonal Communication Motivation of Personnel Problem & Decisional Analysis

Attachment C Current Shift Technical Advisor Training Pronram Command Responsibilities & Limits Stress

• Human Behavior Responsibilities for Safe Operation & Shutdown Cquipment Outages & Clearance ?roceduras Use of Procedures Plant Modifications Shift Relief Turnover & Manning Containment Access Mafncaining Cogni:ance of Plant Status Physical Security Control Room Access Radlo; 3 1 cal Control Instructions Radiological Emergency Plan Code of Federal Regulations (appropriate sections)

Phase III -

Plant Systems Phase Q weeks)

Plant Systems training 61 compasses essential nuclear steam supply, secondary and emergency systems. The student will learn the gancral description of the system, instrumentation and controls, interconnections with other systems, operational li=1ts and basic operation. The provisions of Technical Specifications (including bases) will be stressed. Integrated plant operations will also be introduced. A tentative list of systems to be included in this program is given below.

The final list of systens will be completed after consultation with the Peach 3ottom training staff.

Existing Peach 3occom training materials will be used to the extent possibic.

Emergency Core Cooling Smergency Cooling Water Emergency Elcetrical Power, AC & DC itcaccor Jrotection Reactor Coolant Reactor Coolant Inventory & Chemistry Control C o n ta i ntae n t System Closed Cooling Water Nuclear Instrumentation Non-Nuclear Instrumentation Reactor Control Containment !!ydrogen :tonitoring L Control Radioactive Waste Disposal (Liquid, Cas, SoliJ)

Emergoney Control Air Condensate & Main Feedwater

Attachment C Current Shift Technical Advisor Training Program Auxiliary Feedwater Reactor Vessel Water Level Control Main Steam Status Monitoring Seismic Monitoring lesidual Heat Removal Radiation Monitoring Main Turbine & Generator Phase IV - Accident Analysis Phase (3 weeks)

The objective of this portion of the program is to prepare the STA to perfor= the accident assessment function. The methodology of accident analysis will be presented. Indications and the tasponse of the plant to various accidents described in ve,sdor accident analyses and the Final Safety Analysis Report will also be discussed. Transients of moderate. frequency and infrequent and limiting faults will be covered. Course materials for this portion of the program will be developed from plant specific materials and provided each student.

Phase V - 393 Simulator ?hase (3 ~ a c t: o )

3 oiling Vater Reactor Si.sulator Training is an essential supplement to the classroon instruction and enhances the student's knowledge of the material covered during all four classroom phases.

Training on a full scale boiling water reactor simulator is available from General Physics Corporation utilizing the facilities. The program includes four hours of classroom instruction and four hours of " hands on" simulator training each d ny. Students will bccone familiari:ed with normal )lant operations during Week 1.

Week 2 features transients of moderate frequency.

During Week 3, inf equent and lisiting faults will be explainad with special enahasis on the lessons learned froc Three Mile Island.

The recommended class size for the SW2 simula or training is 3-4 personnel. As we anticipate that 6 students will attend, the training will be given to students in two groups during a three-week period, each 3:aup using the simulator f our hours per day.

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Telephore 2l7166 90!l l Twx NO - 3 00-d'3 9 YANKEE ATOMIC ELECTRIC COMPANY s.l.1.1 7 WC 80-31

&) 20 Turnodeo Road Westborough. Massachusetts C1581

.Ya_uxas Cctober 7, 1980 Mr. Darrell G. Eisenhut, Director Divisica of Licensing United States Nuclear Regulatory 00==1ssion Washington, DC 20555

Subject:

Cc=ments on *Prelisinary Clarification of TMI Acti0n Plan Requirements" feca Oarrell G. Eisenhut to All Licensees dated September 5, 1980 Dear St..

Tanxes Atomic Electric Company feels it necessary to cc==ent en the suoject accument, because as a sajor revisien to previcusly sandated ceccit=ents it represents a. significant increase in werk and an invalidatien of some werx already completed. In additon, since it has been issued for apparently no other reason than as a ra 0het en existing requirements, it constitutes a big " surprise".

The subject docu=ent Ocntains =any .eignificant enanges to and extensi0ns of the existing Category "3" requirements originally detailec in NUREG-0573.

We are perplexec ey these sajor changes to previ us NRC sandated concit=ents at a point in ti=e wnen consicerable ti=e and resources have teen expencec in meeting these ce= sit =ents. Even scre unsettling is the fact that these

" clarifications" emerge only three =ccths prior to the originally =ancated ccmplet10n cate. Past expenditures of time and resources to cceplete an item labeled no longer acceptacle in the sucject document becomes a vaste of manpcwer wnich we can ill afford at this time.

Specific ces=ents en individual "clarificatiens"is the subject document are attached to this letter. However, seme general 00 =ents are offered as follows:

. More and more frequently licensee action is required based en draft docu=ents such as the subject document. This practice shoulc be avoicec because it eliminates the balance provided by industry review and 00==ent and invariably leacs to confusion, duplicatien of e f fort , anc uasted ar.pouer resources.

. Spreading out the recuired submittals over an appropriate perice of co\

ti=e will remove an unnecessary burden from botn the licensees and tne staff. Uncer the current senedule, ne NRC will te unab le to ,

review the rass of =aterial that will arrive January 1, 1981. ' 4 7(

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. All of the Cet0ber 1,1980 su:mittal dates f0und in the su ject cecu=ent snould Oe Onangec to .-eflect the issuance date of One final clarification accu =ents. (CD

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United States Nuclear Regulatory Commission Octooer 7, 1980 Mr. Carrell G. Eisennut, Directer Page 2 l

. In many cases, information requested in the document has already been subt.itted as ; art of an Owners Group effort. Rather than akin 1 a blanket request for new information, an effort should be made to identify the additional information necessary and request only that information.

. Any reference to Regulatcry Guide 197, Revision 2, should be deleted because the status of that document is questionable. We deuct that suppliers can ever meet the new require =ents suggested in tne draft.

. Guidance for writing new technical specifications in the form of revised standarcized technical specification (STS) pages is not very helpful to acn-STS plants. Suen guidance should be in the form cf requirements or objectives whien leave the licensee flexibility of deciding hch to meet the requirements.

. A letter tc Chairman John Ahearne, dated August la,1980 (attacned),

stated that the behavice of iodine released by th) fuel durius postulated nuclear accicents is not correctly described by existing NRC medels and that iodine releases calculated by these eccels are overestimated by =cre than a facter of 103 *his discrepancy should te i= mediately resolved in light ef its effect on:

1. Habitability analyses (Control 3com, Technical Support Center)

2. Iodine Monitoring Requirements 3 Requirements for Pocassium Iodide availability We urge you to carefully censider new requirements to ensure that there actually is a requisite enhancement of plant safety before moving forward with them. It appears t at the staff has devoted excessive attentien to prescribing specific changes instead of estaolisting overall requirements. cr objectives, and eraluating these new objectives in the context of real and meaningful improvements. Additionally, sufficient time sust be provided to licensees for careful design and thougntful review of sedificacices. A convenient =eans to address this area of deficiency is to release drafts with sufficient time for ec= ment. A hurried review, vnile preferable to no review, is never in the best interest of quality wort.

We hcpe tnat o"ur ccaments in this letter and its attachments will be helpful to you in precucing a final version of the subject docu=ent. Should you have any questions resarding our comments, please fsel free to centact us.

Very truly yours, -

l YANKII ATCMIC ILECTRIC COMPANY l

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D. W. Idwards Oirector of Operational Projects l

Ccaments on Scecific Items of the Preliminary Clarification I:.E.3.2 Containment !sciatien-Decendability Ccement:

1) Reducing the ;ressurs setpoint for entainment isolation may not be compatible with attempts to reduce Scram /SRV system challenges as described in Item II.K.316 of ene five additional TMI related requirements.

2) The requirements that purge valves to be sealed closed and checked every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is unreasonable, one or the other should be acequate.

I.A.1.3 Shift Manning Ccmment:

1) Overtime restrictions snould be provided as guidance, not as requirements. Deviation from *,hese restrictions snould be all0wed as authorized oy upper plant ?*nagement.

2) Flexibility snculd be given to plant management to allow resolution of conflicts with existing labor agreements created by ene new requirements.

3) Current Stancardi:ed Technical Specifications do eat seem to be consistent with the clarification of Item I. A.l.3 I!.3.2 Plant Shielding Clement:

1 The clarification should state that only direct radiation doses (anine) need be censicered for this item. A statement was made to this effect a*

the meeting of Septemcer 22, 1980.

2) Implementation dates for plant modifications should be consistent with the recent N3C order regarding environmental quclification of equipment.

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I:.9.3 Post-Accident Sameling Comment:

This item has been discussed with NRC several times with respect to l obtaining a chlorice sample, and no technical basis has ever been given '

for requiring sucn a sample on a short term basis. The chicrice samoling requirement cannot be met through dilution sampling aoproac due to the nature of the sampling tect.11gue. Even if a sample coald be taken 1:

i would be of little or no value in a post-accident situation. An alternate means of determining possible caloride intrusion to the primary l system would be to sample secondary systems which could contaminate the  ;

prinary system.

II.D.1 Valve Testing Recuirements 00mment:

1) The possibility of using the required valve testing facility to test valves to ATWS requirements has been discussed with the NRC several times. At the last discussion, the NRC agreed that ATWS testing coulc not be acccmplisned at this facility due to senecular restriction. It was also agreed that the final " clarification" of the TMI action plan requirements would be revised to eliminate AT*43 testing fecm the program. To impose ATWS testing requirements st this point in the facility's construction would mean scrapping much of the existing equipment, and starting over. Major implementation delays wculd result.

2) It is unreasonable to require that the EPRI/3WR test data results and plant specific sutmittals be su mitted on the same date. C0mparison of the test results to plant unique situations will require seversi months.

II.F.1 (Attachment 1) Accident Menitoring, II.F.1 ( Attachment 4) Containment Pressure Monitor and I!.F.1 (Attachment 5) Containment Water Level Meniter Comment:

Current information from the vendce indicate that the current implementation date of January 1,1981 may be impossible to meet due to unavailability of equipment.

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Augus 14, 1983 4

Chair:en John Ahearne '

U.S. Nuclear Regulatory Commission 1717 H Scree:

Washing:on, D.C.10535 Cear Chai. an Ahearne:

We wish to bring to your at:ention a ma::er tha: =ay be a very important devel-opter.: in reac:or safecy analysis. We believe : hat sufficien: evidence has accucula:ed :s show thac :he behavior of iodine 'during nuclear reac:or accidents is not corree:1y described by exis:ing NRC models and Regula:ery Guides. Iodine volatili:y is grossly overescisa:ed by these models for accid 2nts in which sub-stantial amoun:s of va:er are present, and escape of iodine .o the environmen:

vill be ex:re=ely small (as 1: was at Three Mile Island) as long as reasonatia containnen: integri:y is also maintained. As a consequence, the risk to the general public presented by iodine is lower chan estima:ed, perhaps by ceders of magni:ude.

Our concern vi:h :his issue originated vi:h our involvement in the several Technical S:aff Analyses for the President's Commission on the Acciden: ac Three Mile !sland. The mechanism for the behavior of iodine : hat we propose here was derived from : hose analyses, from further examina: ion of experi=en:a1 and

heore:1 cal studies involving :he chesis:ry of iodine and casius fission pro-due:s in ligh: va:er reactor fuel and systems, and from :he observed behavice of iodine subsequent :o fuel failures during acciden:s and incidents a: other reac-

or si:es. We believe :ha: :he explana: ion presented here vill change :he pre-sent concepts of :he ha:ards involved during and subsequent to reac:or accidents and, therefore, will require a cri:ical reexamina: ion of how :hese ha:ards and risks are calculated, and :he cri:eria :o which engineered safeguards are designed and ins:alled.

A1:heugh :he Three Mile Island (TMI) reac:or core inven:ories of xenon-133 and iodine-131 vere co= parable, between 2.4 and 13 million curies of xenen escaped

s :he environ =ent during the accidene, while only 13 :o 18 curies of iodine similarly escaped! This grea: dispari:7 was identified as a mac:er of crucial impor:ance early in the investigation by :he ?residen:'s Co= mission, and an effor: was =ade to find the explanation. Ic was clear tha: we could no: claim

o understand :he acciden: until this discrepancy (a fac:or of 105 :o 10 )6 was explained satisfactorily. Further, 1: was recogni:ed :ha: :he physical snd chemical condi:icns during :he accidenc 4: O!! =ay not have been unique. (Ve note : hat, generally, radiciodine is :he con: rolling fission produe: species vi:h respec: :o si:e safe:y analysis as well as :he design and operation of certain engineered safeguards. )

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Chairman J. Ahearne Augus: 14, 1880 The explanation for the very low escape of iodine :ha: developed during :he invas:1ga:1on by the President's Consission was :ha:, as :he :emperature of :he core increased, iodine diffused out of the fuel rods through the failed cladding and vaporized. The iodine escaping, if not already in the iodide form, then encountered a chemically reducing environment which converted i: :o iodide. The iodide subsequently wen: in:o solu: ion as iodide ion when 1: con:se:ed wa:er.

I: vas recognized : hat addt:ional experimental work was needed :o provide a quan:1:stive descrip:fon of the iodine behavior. Neve rtheless , :his explanation accoun:ed for :he much smaller escape of iodine : hat was observed a: IMI com-pared :o :he amoun: predic:ed :o escape if elemental iodide had been present, as is assu=ed in the Regulatory Guides.

'44 believe tha: :his descrip:fon can be s:reng:hened and made more defini:ive.

A1: hough :he presen: ta:a are no: absolutely conclusive, we believe :ha: iodine acerged from the fuel as casium iodide, already reduced :o iodide. The reac:or systa= environ =en: : hen sus:ained :his chemical s: ate. yurther ore, i: vould have converted c:her iodine species, should they have been presen:, to iodide.

Cesium iodide would be expec:ed :o condense or "pla:e-out" when 1: reached me:al surfaces 4: tempera:ures a: or belev 400 :o 500*C, and i: vould finally en:er in:o solution as iodide ion as soon as water or condensing steam was encoun-

ered. The reactions of iodine species in water, and :he fac: :ha: iodide ion is the domi a.an: species, ensure :ha: iodine volatili:y will be very small (compared :o :ha: implied by :he Regula:ory Ceides, for example). A reac:1cn causing oxida: ion of iodide would be necessary to increase :he vola:ili:y of iodine. Addi:1onal experi= ental work is required to provide a quan:1:a:ive descrip: ion of iodine behavior, bu: :his quali:stive pic:ure is consis:en: vi:h

he r=all escape of iodine observed in a number of inciden:s when wa:er was pre-sen:, such as at "M1.

This mechanism is supported by the folleving observations, as wel; as by

=easure=ents made a: IMI:

1. Iodine and cesium are released congruently from ?JR leakers during power

ansients (:he iodine spiking phenomenon).

2. Ther=odynamic calculations performed a several sites indica:e :ha: Cs! is the s:able form of iodine in 1*4 R fuel. Further, che fission yield of cesium is larger :han tha: of iodine, and cesium is always presen: in grea: (abou:

tenfold) excess over iodine.

3. Irradiated fuel has been caused to fail in experi=ents perforced under si=u-la:ed acciden: conditions, and the iodine released is recovered predo=1-nantly as CsI ra:her :han as molecular I *2

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Chairman J. Ahearne August 14, 1980 4

4 The chemis: y of iodine is such'tha , if va:er is accessi; ole, iodine vill

, interac: wi:h the water so :ha: 1:s concentra:Lon in :he gas phase vill be such smaller chan 1:s goncen::ation in the va:er.

5. In other incidents that have led to the des: ue:1on of fuel in water sys: ems ,

(NRX, Sper:-1, Snap::an-3, SL-1, MTR, ORR, and yRTR), we unders:and :ha: a much smaller amoun: of iodine escaped from :he sys:e=s :han would be pro-jected by :he existing models. Data are hard :o eces by for many of :hese accidents and experimen:s, and our inves:1gation is con:inuing. In marked con::ast, a large frac: ion (20,000 curies) of :he iodine escaped to :he environmen: during the Windscale accident, which occurred under oxidicing c=ndi:1ons and in the absence of water.

The significance of :his mechanism for iodine escape and :: anspor can hardly be j overs =phasiced. We asserc tha: the unexpec:edly lov release of radiciodine in '

he IMI-2 accident is now unders: cod and can be generaliced :o other postulated '

acciden:s and :o other designs of wa:er reac: ors. We believe :ha: an accident involving ho: fuel and a va:e or steam-va:ar environment vill have :he sa=e con: rolling chemical condi: ions as did :he 21I-2 core and primary system. The iodine will emerge at CaI (ard possibly some other iodidas) and enter into the solucion as soon as ve: steam or va:er is encoun:ered. I: vill persist in solu-cien as non-vola:ile iodide ion as long as oxidicing condi: ions do no: prevail.

A1: hough we feel :ha: :he evidence is sufficiently s::eng to justify :his le::er, i: is i=por:an: :o qualify our posi:1on. Iodine chemis: y is very complex, and defini:1vn experimental and analytical s:udies of iodine behavior during and following loss-of coolant acciden s are lacking. Nonetheless, 1: is

• clear :ha: :he behavior prcjec:ed from :he existing Regula ory Guides is wrong.

The curren: NRC assumption, : hat elemental iodine 'is :he chemical form of :he radiatodine released, is regarded as a conserva: ism, ,bu in this case :he assu=ption of a wrong chemical form sus be regarded as an error which has com-pounding effec:s.

If, af ter due consideration, the NRC is satisfied that our descrie:1on of iodine behavior is valid, we recon =eed that an urgen: study and assessmen: be made of all available informa:icn, and appropriate actions be undertaken. Wi:h due respect we point ou: four consequences should our posi:1on be correct:

1. The frequen:1y quoted fission ; cdue: escape assu=ptions (!:om !!O-14 344 in 1962 to the more recen: Regula:ory Guides 1.3 and 1,4, and :he Reac:or Safety Study, VASH-1400) should be reexamined. The presen: assump: ions I grossly oversca:e iodine release from a reac:or si:e in many :ypes of less-of-coolan: accident, and safety criteria based on these assumptions should be reevaluated.

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Chairman J. Ahearne Augus: 14, 1 C 20 4

2. Be dispersal of radiciodine in 'the biosphere say no longer domina:e and control consideration of acciden:s and the design of safety sys:e=s.

3. Many, if not most, accident sequences must be reexamined in detail. Be iodine risk to the general public may, in fac:, be lower :han previously estimated, possibly by orders of ragni:ude. De impact of a redue: ion of iodine risk on the requirements for evacua: ion is particularly i=portant at

his time.

4 De engineered safeguards designed for iodine control should be reexamined to assure effectiveness and optisi:ation for the ac:ual iodine behavior rather than :he behavior currently assumed.

' Finally, we realize that a major revision of NRC assumptions rela:ive to acci-i dent analyses, dose calculations, and design of safeguards shculd not :ake place without an adequa:e base of technology from boch expert =ent and :heory, and especially until the Commission 1:self is convinced : hat it is appropriate :

accept a revised physical and cheetcal description of iodine :ranspor: from fuel

o :he environment. Cu the other hand, the impact of wrong assu=ptions is so serious : hat an intensive effort should be made :o establish the facts.

We are ready :o of fer more de: ailed informa:1on or further assis:ance should :he NRC reques: 1:. We will be pleased :o brief :he MRC staf f or any review co==1:-

aes you =ay appoin:.

Sincerely, W. R. Str.t:en.

Los Alamos Scientific Mboratory M ,

A. ?. Malinauskas Oak Ridge National Labora:ory

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3. O. Campbell Cak Ridge National Laboratory 4

cc: G. V. Cunningham, DCE-4* ASH

3. M. Kerr, IASL H. Postsa, CRNL e w-- -w-- ,- , - - ,, , - . , , - ~ , , _ - - , - - -

AMERICAN ELECTRIC POWER Sen; ice Corporation ggp 2 3 road cay. Xe:c Tark.3. T. ;0004 1212) 440 9000 October 3,1980 CG-42 Mr. O. G. Eisennut Director Divisien of Licensing Office of Nuclear Reactor Regulatien U.S. Nuclear Regulatory Comission Phillips Su11 ding 7920 Norfolk Avenue Sethesca, Maryland 20014 Cear Mr. Eisennut:

CCMMENTS CN NRC POSITICN CN TTEM I.C.1 The purpose af this letter is to provide the Westingneuse Cwners Group cements en the dnft NRC Position en Itam I.C.1 entitled, %idance for ce Evaluation and Deveicement of Procedures for Transients and Acddents per d the Septemcer 5, 1980 NRC letter on " Preliminary Clarification of TMI Action Plan Requirements *.

The 'Aestinghouse Owners Group believes that all of the requirements regarding tne action items identified in NUREG-0573, Item 2.1.9 for Items I.C.l(a)2, Inadecuate Cort Cccling, and I.C.l(a)3, Reanalysis of Transients and Accidents, have been comolied witn as a result of price or ongoing Cwners Group acdvities.

The draft NRC position I.C.1 is the first indication fmm the NRC that taese recuirements have not been c:molied with. 'de de not feel that sucn a position is warnntad at this time since all recuested informa* ion nas either been sub-mitted ts the NRC or will snartly be submittad. The infor-r.ation relating to the dnft pcsition i* ss fellcws: ,

1) Mccified Wesdnuncuse Guidelines - Emergency Openting Instructions have caen sucmittaa to .ne NRC wnica shculc satisfy this item. Sucmittals regarding guidelines were provided :s the NRC in lettars of June 29, Sectamcer 11 and 23, Oc:ccer 16 and 31, Novemoer 2 and 5, Cecemcer 23, 1979, .

January 23, and July 15,1980. Westingncuse Cwners Gecuo Guidelines E-0, E-1, and E-3 were subsequently reviewec and approved by the NRC. The loss of sec:ndary c:clant guideline, E-2, was submi::ed :s the NRC wita l

ne other guidelines but nas not as yet been reviewed by the NRC saff.

2) Innsient and Acddent Analysis Item 2.1.9.c of NUREG-C573 - WCAP-9691 was sucmt: tac c :ne NRC on Maren 31, 1980 wnien snoula sadsfy :nis 1.am. ~he WCAP ::ntains even rtes for :ne ma,fer ac:idents and an evaluation of :ne covenge of the event secuences in :ne Westingneuse Owners Gecuo EDI's. Up :s this :cin; in d=e, no feecback has been

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i Mr. D. G. Eisenhut provided on the status of this reocrt other than the dnft NRC cosition.

We believe it would be are acercoriata to review this submit a1 in order to assess c mplianca with the 2.1.9.c recuirsments.

3) Inadecuate Cort Cooling (ICC) - A preliminary guideline and analysis on Inacequata Cort Cociing were submitted to the NRC staff in letter of October 30. 1979. Westingnouse has subsecuently submittad three WCAP's wnica provide a re ccmcrthensive analyses of ICC events. These WCAP's are WCAP-9744, " Lass of Feedwater. Induced Loss of Coolant Analysis Report", WCAP-9753, "Inadecuate Core Coaling Studies of Scenarios with Feecwater Available. Using the NOTRUMP Comouter Code", and the report attacnec to letter NS-TMA-2279 dated July 24, 1980, " Inadequate Cort Cooling Studies of Scanarios with Feedwater.Available for UHI Plants j Using the NOTRUMP Camputer Code *.

Based on these are comorthensive analyses, the Westingneuse Owners Grouc is in

ne precass of developing revised reference inadequate cort c: cling instructions ts cover multiple failure events of small loss of coolant accidents without hign head safety injection and ccmDlete loss of fetcwater. The ICC guidelines ar*

scheduled to be summitted to the NRC by the end of October,1980.

In the NRC positicn I.C.1, several points are raised wnich are addressad by the folicwing c:mments:

1) Justificaticn for the approach taken in develcaing diagnostic guicance has been provided for each guideline that has been submittad to tne NRC.

Tnis is escecially true for the emergency guideline diagnostics con-tained in E-0, *Inunediate Actions and Diagnostics", wnich was estansively reviewed by the NRC prior ts approval.

2) There are several instances wnere the emergency guidelines address the availability of systams and corrective / alternative actions that should ce perforned to mitigata the event shculd syttams or c:moonents fail. Scme examoles art; 1) giving the operator several alternatives in order of preference to deprsssurize the RCS in tne Staam Genentor Tube Ructure Guideline, E-3; 2) establishing power sources .s equicment early in ne guidelines wners the coarstar must per'orm scme action or actions to make the acuipment available wnen needec; 3) crovidtng contingency actions for One swit:nover to ECC3 recirtulation.

3) Multiple Failures have been addrsssed ':y WCAP-9691, *NUREG-0578 2.l.9.c Transient and Accident Analysis", in event trees for costulated emergency situations. This resulted in some changes to the EDI guideTines wnica were inc:rporated in Revision 2 submitted ts the NRC July 15, 1920.

Prior to making additional adifications to tne guicelines it is rec:m-mended that better clarification of r:11abili y goals is needed along with an evaluation of ce overall risk.

In regard to ce examoles of multiple failure given in the NRC ;osition, Items 2 and 3 (ICC) are being adcressed by guicelines is be sucmitted in Cet:ber,1980 and Item 4 (ATWS) is being adcrassed ;r cadurally in plant scecific procacures as recuirec by the NRC. Item 1 (Multiole Steam Generator Tuce Ruoturts) is acortssed in :ne NRC evaluation of ne I-2 guiceline in :nat it statas nat an evaluation will be mace to detarmine i

Mr. O. G. Itsenhut the accactability of the limitation of one tube ructun in existing guide-lines. We have teen given no evidence tnat tnis avalu:. tion had c:ncluded that ne curnnt range of acclicability of the E-3 guideline is unaccactable.

Itam 5 (Ocerator Irrors) is acdressed by :ne guidelines in tnat they attamot to minimize the possibility of coentar error by minimi:ing recuired opera-

.sr actions in the guideline. Fur-hermars, in regart to ocers:cr errar, tne perator is instructed in the guidelines to continuously monitor imcor-tant parameters subsequent := per#0rming operator actions. Unantici:atad plant response would alert the acerator tnat an operator error has occurred.

4) As stated previously, analyses and guidelines have been and soon will be ,

submitted to NRC on Inadequate Care Cooling. We believe 21s should ccm-plete all the comitments associated with Inadequate Core Cooling to date.

In c:nclusion, the Westinghouse Owners Group in reviewing the NRC Position I.C.1 finds no new scecific recuirsments beyond those recedures or changes t cre-cedures wnich have alreacy been comoleted or will be furnisned in the near future. Thertfart, unless additional clarification is provided by the NRC which resul s in new work being defined in :ne procedures area, we feel that most of the itams mentioned in the NRC position have been or will be addressed. No additional work beyond that already c:mittad is planned. If one NRC Staff would find it useful, the Westingneuse Owners 3rcuo would be willing to meet with the staff in order to discuss the NRC position and our c:me .ts in mort detail.

Very truly yours, Robert W. Ju ens Chaiman Westinghouse Cwners 3roup

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PUBLIC '

SERVICE INDIANA occeber 6, 1980 S. W. Shielcs Senior Vics P oscent .

Nue:eer Oimsen i l

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1 W. Oarrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regv.lation  ;

6 S. Nuclear Regulatory Cc= mission  !

'4ashington, D. C. 20555  !

Dear Mr. Eisenhut:

The in:ent of :his letter is to provide writ:en ecmcats with respect to

he report, Letter to All Licensees of Oeerating Plants and Acolicants for Ocerating Licenses and dolders of Constructien Permits, dated September 5, 1980. '4e understand that the report is considered to be clarification of some of :he 2C Action Plan requirements, but is not yet final. '4ri::en c:=ments were requested during September 21 Region III neering in Chicago.

I: is also our understanding that :he ec==ents are needed on a :i=ely basis for NRC staff consideration prior :o going before the Comissioners.

As you kncv, :ha requirements and rece==endationo which have been for=ulated

hrough studies of 2C have found :he form of various reports and letters, and an understanding of their intricacies has been difficult for most involved.

Pub:1c Service Ceepany of I= diana, Inc. (PSI) does not have the sa=e schedule constraints as nany other nuclear utilities, as our fir'st two units are not scheduled to be ecmplete u=:11 1986 and 1987. *4e believe :he cempilatien of CE related requirements in WAECs 0660 and 0694 (NRC Action Plan as a Result of the TMI-2 Accident and "MI-Related Recuirements for New Ocerating Licenses),

and now :he September 5 letter, is a good practice which should be continued.

Basically ve feel the report represents some good thinking. Frem what we heard at the September 21 meeting, plus frem others in the industry, there has been difficulty on the part of licensees in meeting seme CC-2 Action Plan subcasks ca schedule. The licensees have been at work on the 1: ems, and have encountert. problems in procure =ent and construction in scme cases. l O\  !

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Mr. Darrell G. Eisenhut October 6, 1980

, The September 5 document has accordingly factored this, plus some other

{ problems into new schedules for some of these tasks (as examples, the instanation deadlines for reactor coolant system vents has been =cved back 12 months, to January 1,1982; and the completior. date for emergency support facilities has been moved back 15 months, to April 1,1982). Also we note that the NRC staff has further defined what is needed to see same of the tasks, and new activities have been proposed for some subcasks.

!bere were additionany some " disjoints" which have been recognized, the specifics of which were pointed out in the meeting. Generany, they fell into the following types of areas:

o Some licensees have already completed designs and installations.

New requirements for them might in effect petlize them for making the previously established schedules, as retrofits may be required.

o The NRC apparently has not been able to support the licensee efforts with respect to submit al review. In scue cases, the required dates for installation may occur before the NRC has initiated review.

o Scue of the recommendations which were used as input to the IMI-2 Action Plan sounded very good; however, there is now concern whether the necessary instruments and systems can be adequately developed in support of the required schedules. (One example concerns the require-ment for the instanation of an instrumentation system which gives the control room operator a clear, unanbiguous indication of inadequate core cooling; Task II.F.2. Both the completion of the development program and the required installation have the same scheduled date:

January 1, 1982. PSI has doubts whether such an instrument has been or can be adequately defined) .

I o For cperating plants, implementation schedules for activities which would require najor reactor shutdowns should taka into account individual plant refueling schedules.

o In some cases, concern was expressed over addi$ional proposed act-ivities that were set accompanied by relaxed schedules. (We recognize that schedule adjustments are not always necessary) .

We have one general concern: We feel that the implementation schedule for some of the requiremen*s is very tight for plants approaching opurating license or in operation. We appreciate the pressure that the NRC is under to get the TMI lessons implemented. Nevertheless, we recomend that a re-eyse hmtion of the required schedules be made in order to fully define the required systems and the1: respective contributions to overall plant safety.

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i Mr. Darrell G. Eisenhut October 6, 1980 Again, please continue the practica'of acctmulating the current status of requirements into one doc;maant. We would be pleased to discuss this or any other of our comets with you if you wish.

Sin ly, j

! S. W. Shields RSW:gb e

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o Jereercommes poser a ugna campany JCP&L G P fJ -

Momsamun, New Jamey W5ED (201)45EHI200

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Septemmer 25, 1980 Mr. Dar rell G. Eisenhut Director, Division of Licenslag .

U.S. Nuclear Regulatory Caummission >

Washington, D. C. 20555 -

Dear Mr. Eisenhut:

Subjec+: TMl2 Rela ec Recuiremen s for :erating Reactors INURIG 0660)

Oyster. Orees Nuclear Genera *ing Station Decket %s. f0-2:9 in response to your letter ca ec September 5,1980, JCP&L has rev ~ emed al1 of Oyster Creek's TM t lessons learned activ ities against the clarified

.;,n .

, requirements of that letter. A= a result of that review, we are requesting that the implementation cate of' scue items be attended.

Attachment A and S to this let*er prov ide alternative schedules and justification for two items: II.K.3.14, Isolaticm of the Isolation Condense s on High Radiation in Vent and li .K.3.27, Commacn Reference Level for Vessel ' eve! _

I nstrum entation. Our position on 'hese two i+sms was originally addressed- in our June 23, 1980 istter written in response to your May 7,1980 letter.

Also, s i n ce a o I ant snut:own would be required to ccep i ete i*em II .K.3.19, Interlock on Recircul a?ien Pum: Lecp , we woul d propose com:l et! g that item during a Spring 1951 snu a:wn schecul ec ter the purpose of ceco: eti..g cther action plan Itees. We fee l this is justified since JCP&L has at: ea:y imp lemented coeprehensive adininis rative controls to assure at l ess tic recircul ation loops are open at al! times. These administrative controis l Include: establishing ttils as a Saf ety Limit in Oyster Creek's Technical .

Specifications, procedure changes, hinged covers over all recirculation Icep isolation valves and consolcuous warning signs on each cover.

! :urther, the Fecera! Env iren=ertal c ecti on Agency's Na**o a!

ol l c* ion D ischarge Permir ter Cys e ~. ee, ;. en::i s in Section 9.:.5, : sr. ec snu cowns durIng December thecugh var:n . This : eaves a very beief per:c: t:r planning and preparation fcr tne accificaTien, including provisions for aceauz e ALAftA review.

We al so f eel that a shurco=n before January 1981 would no' f aci: 1 z e me c:rapietion of other items, bu wo:lc in fac* divee* resources fran tne i u::s

'nr* are required to be c:r s le*ec :y .a9us y * , *?B*. r,

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W . Darreli G. Eisenhut September 26, 1980 JCP&L cannot yet connit to c=mpleting the revised requirements of " tem l i .E.4.2.3, Containment isolation .Desendabi lity, until we have had a chance to review the additional NRC guidance on requirements for three (3) c*vs se isota-lon signals and essential vs. ncnessential systems.

I f you should have any questions regarding the response, please s. oct P . James Knubel (201-455-8753) of my staff.

Very truly yours, j

' Ivan R. Fi ntro 4, J

. i ce 2 reside 9

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n ATTACHMENT A

!SO*./.TI:N 7 TrE ISCLATICN CONCENSERS ON HIGH RADI ATICN IN VENT (NUREG 0660, II K.3.14)

MtO REQUIREMDfT Modify the isolation condenser logic so that the isolation concensers aro isolated from tne reactor igen the receipt of a high radiation signWI in the c acscheric ven* soalter insteed of the steam line monitor. The goal at nis --

1 modificrrice is to increase the availability of the isolation condensers curing - '

an acciderrt by isolating thee only if there is radiation being released item the picat through the shell side vents to the atmosphere. Such a release condd sily "

oce:.r if *%ere we e a tube leek. Modifications shou ld be conp lete by Jartary 1, 1931.

OYSTER CREEK's RtESENT DESIGN The isolation . condenser isolation l og ic is scme whart d i f ferent frtan - .a-desc-Ited for the *ysical BWR. A vent line radiation monitor does exist wh*cn ata ms :n the con rol room, however Oyster Creek coes not have a sten !!ne radiation monito weich is used to isolate the reactor from the condensers.

OTSTER ctEEK PostT108 JCP&L agrees with the NRC's goals of increasing the availabil ft.y of the '

Isolation aandensers, post accident. They are passive ~ systems and are welI cuited for Iceg term core cooling. JCP&L's concern is that if this modi #Teatice le ret properly designed, it wil have an adverse effect on availability. To oc this .io: properly we cannot simply rowire the existing vent line ratiattom monitor alarm into the isolation circuit. In our judgemeat, the esisting monitor would alara, even if there were no tube failure sinply because of %e, ectronely high radiation levels expected in the area of the ;.onitors from -he

• icola e: conce ser's steen & condensate piping. Furthermore, it is imp acti:a' ic s-ield -he prese .t detectors suf f iciently to ensure that their alarm'ng imo1ies a 'uce rupture.

JCP&L feels that a modification to this system which achiews ne dos ree results cannot be completed by- January 1, 1981. At this time he required radiation monitoring scheme has not been specifIed and obvioust y -he  :.===

procurement lead times are unknown. We propose to submit a design 1er t:Is -

modif ication by January 1,19 81 and at that time wi l l also incisde ar le= lemer: a-ion schacu!e based upon vender sue:Iied delive y dates.

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,. ATTACHENT B 006040N REFERDd2 LEVEL FCR VESSEL 1.EVEL INITRUMENTAT*ON (NUREG 0660, 11. K.3.27)

WIC itEDUIR9e(T Modify all reacter vessel water level instrumen ation s: ther tasy have a common reference level. Complets by January 1,1981.

OTSTER CIEEK's POSITlWI in general'we agree with th concept of having a commeon reference level for alI reoc or vessel level Instrumentation. Most pe sons no- involved with con rel roam operations find the present system cf l eve l "n strum e r.ta-lon cumberseme mc41'f ficult to ecmprehend. Control r:om opers' ors anc s-if t susev is:rs ,nc have been werking wi-n tne present sys a- f:r a ucce of years, ho wev er , are quite comfortab le with it and adept at usicg it. JC:&L i s concerned t:s? an abrupt change fecun the present sys se may Tave a confusing

. ette:t ec c=nsequen'ly be the cause of operational prot l ens. In additic:, a substan-ist aoninistrative task is associated with t-is e'. terr-lon. Relabeling the ' aces =f the Indicators sid recorders is relatively sin:l e :ompared to t 1e ta st of dentifying and revising all logs and opera-Icns, emergency and surveillance procedures which use the present system of level instrumentaticm.

JCP&L proposes time following alternate schedul' e . By Oc ober,1980, all level indica *oris and recorders will have vieir faces alter,et shoeing two scales.

Orne will be the present aarkings and the other .will be referencad t= the tco of the a:tive fuel . We will a.z; ioentify all documents te.g. legs and procedures) which wil! seed to be revised e a result of changing m refrence point. The period between October 1980 and tne 1981 refueling outage wil 1 :e a transifion period curing which the operators will becone accustoned o the nem scales and changes m alI documentation viIi be drafted. The eliminstI:n of the o1d seales and charges to alI affacted procedures, logs, etc wI' I oc=ur during the 1981 se-a;e . l *s felt mat an cutage period is the bes -f ee 'c m i s trans !-!on secause level instrumentation in the normal operating ranges is nor usec and therefers the opportunities for coerator misjudgement a c minimi:ed. The sp-ing outage w i 2 I al so prov ide a b Iock of time during wnich me required 1 aining sessions can be conducted.

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