ML17138B518

Confirms 801027-31 for NRC Visit to Prepare Preliminary Evaluation of Control Room W/Regard to Listed Items.Requests That Util Perform Preliminary Human Factors Assessment of Control Room.Assessment Info Encl

ML17138B518
Person / Time
Site:	Susquehanna
Issue date:	10/01/1980
From:	Stark R Office of Nuclear Reactor Regulation
To:	Curtis N PENNSYLVANIA POWER & LIGHT CO.
References
NUDOCS 8010210324
Download: ML17138B518 (34)

Text

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Qgg)g@ ".Pied BIedeeco DEisenhut, RPurple Ot T 0 19 JYoungblood RMStark MRushbrook PCheck Pennsylvania Power 5 Light ComPany,

,, LRubenstein.

ATTN:

N. M. Curtis, Yice Presidgnt Engineering 8 Construction 2 North Ninth Street

..,, $Fj,ragl,ia Allentown, Pennsylvania 13101...,.',,Ryan.amer...,

DRoss Gentlemen:

RMattson RHartfield, MPA OELD OIE (3) bcc:

SHana'uer JKr amer

VMoore, DTonds.~

.DVas sal 1 o DZiemann PCol 1 ins HFEB Members TERA NRC/PDR L/PDR ACRS (16)

NSIC TIC g-D -3173fF This letter is to confirm the dates,October,27, 1980.through October 31, 1980 for a visit by the NRC staff to your,control room.

The,objective of this visit is to prepare a preliminaryevaluation pf the control room with regard to the following:

1.

The adequacy of information presented,to theooperator to reflect plant status for normal operation, anticipated ppet ationa1 occurrences, and accident conditions; 2.

The groupings of displays and the layout of panel~;

3.

Improvements in the safety monitorjng.ang human factorsenhancement of controls and control displays; 4.

The communication from the controlrooy,tg points, outside, the, control room, such as the on-site Technica]

Support Center.

(This communication link must also be coordinated with new requirements for transmipsioy of.

plant systems data to NRC);

5.

The use of direct rather than derived signals for, the presentation of process and safety ipformation to the operator; 6.

The operability of the plant from the control room with multiple failures, of non-safety-grade and non-seismic systems,.and control room systems; 7.

The adequacy of operating procedures and operator tratl,ning with respect to limitations of instrumentation displays,jn

)he, contro1 room; 8.

The categorization of alarms, with unique definition of safety alarms.

To assist the staff in its evaluation, your, cooperation is requested in the following ways:

1.

Admittance for several people to the control room as needed for the duration of the visit; 2.

Two qualified control room operators oI; sipulator instructors to, assist the staff for the duration of,the visit; P

e OFFICE SURNAME DATE$

RC FORM 318 t9.76) NRCM 0240 927031 AU.S. GOVERNMENT PRINTING OFFICE: 1979'289 369

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Permission to use video and audio, recording equipment and photographic equipment in the control room; 4.

Access to copies of selected emergency procedures used in control room operations; 5.

A briefing and debriefing with, the plapt manager.

We expect to be at the Susquehanna

3. security gate at about, 1:00 p.m.,

on Honday, October 27, 1980.

The visiting team,><))1 consist of members of the HRC staff and its consultants.

Prior to our visit, we request that you perform p preliminary human factors assessment of your control room as ~oted in paragraph I.D.1 of NUREG-0660 to identify significant. hum'an factors and instrumentations problems.

Please forward the resut ts to us prior to our visit.

In

addition, we request that you send us the fol]owing data, photographs, and information along with your prt.liminary assessment:

1.

Photographs (see attachment 1 for suggested format) 2.

Control room drawings (see attachment 2 fpr details) 3.

The following contact's name and phone ngmber:

Applicants site contact to assist fhe,review team 4.

An arrangement for three operators to assist the review, team on Wednesday, October 29, 1980 to perform control/display arrangement review.

5.

An arrangement for a shift complement of,operators to assist,us during procedure walk-throughs on Thursday,,October 30, 1980.

6.

An arrangement for security clearances for the.team.equipment.

Video equipment.

2-35mm cameras and tripod Cassette tape recorders (3)

Film Sound level measuring equipment Light level measuring equipment 7.

A list of abbreviations used on Panel/System(Subsystems 8.

Four copies of four emergency procedures.

OFFICE SURNAME DATE$

NRC FORM 318 (9.761 NRCM 0240 AU.S. GOVERNMENT PRINTING OFFICE. 1979'289'369

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Me would appreciate receiving.y11,Iof.they)oVe,requested,,fyfopmatfoppt, least,,

two weeks prior to our visit so.that'wp ~,haye,t)me, to review,ft.Please

feel free to telephone Mr. S. P. Saba.,the temp.,1epder.,at (301), 492-,8962 ff additional fnformatfon fs requfred,.relatfng to. the, control, roop. vfsft., Youp, cooperation in this important matter fi. yppreciyted..,

Sincerely,

Enclosures:

As Stated cc:

See next page Nchard M. Stark, Project Manager Licensing Branch No.

1 Dfvfsion of Licensing

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I OFFtCE SURNAME BATE$

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gv /8 1 od NRC FORM 318 (9 76) NRCM 8240

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ATTACHMENT NO.

1 (Suggested Format)

Photographs (3 ring black binder)

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1.1 Slides and 8"x10" Photo-enlar ements. of. S ides +ho<<t<<ogryppeg, in the following manner:

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a)

Panorama Shots of the General.Cyst.olRoom~..i.e, Shots taken from the center of.,therqom<<shoping panel

.groups in horizontal series.

(b)

U er Middle, and Lower Close-u Shots, of Pach, Panel, showing panel annunciator/display/console,in vertical series.

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Bf the control room approprtatelyg,Indexed to Identt.fy, Panel number/name with photographs/slides, OFFICE SURNAME DATE$.

NRC FORM 318 (9-76) NRCM 0240 4 U.S GOVERNMENT PRINTING OFFICE: 1979.289p389'

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ATTACHNENT NO.

2 Drawings (Reduced for control rqqp,,qevi,ey,usg).

1.

ELEVATIONS: of all control(dipplygpyyel s,,idgpti frigg,, panel,sg, systems/sub-systems/groups/sub;groups, ay)fngividual,components.

These drawings shall include the.pyocepscgpputet'pd other control.

panels associated with the control, y,oops/.The.,control room,iq. de fined to include all back panels,...

2.

PRANS:

shall include (a) a F'foor P'lan to,identify all spal:es,.work, stations and equipment.. All panels shall be fdentjfied,by game an'rehd.

number, (b) a Reflected. Ceil.in Plan to identify, 4ocation of

"'ighting systems available to thy operator during normal and emergency operating conditions.,

OFFICE SURNAME DATE/..

NRC FORM 318 (9'76) NRCM 0240 4 V.S. GOVERNMENT PRINTING OFFICEI 1979 289 389

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Mr. Norman W. Curtis Yice President

- Engineering and Construction Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18'l01 CC:

llr. Earle M. Mead Project Engineering Manager Pennsylvania Power

& Light Company 2 North Ninth Street Al 1 entown, Pennsyl vani a 18101 Jay Silberg, Esq.

Shaw, Pittman, Potts Trowbridge 1800 M Street, N.

W.

Washington, D. C.

20036 Mr. William E. Barberich, Nucleal licensing Group Supervisor Pennsylvania Power 8 Light Company 2 North Ninth Street Al 1 entown, Pennsyl vani a 18101 Edward M. Nagel, Esquire General Counsel and Secretary Pennsylvania Power 8 Light Company

~2 North Ninth Street Al 1 entown, Pennsyl vani a 18101 Bryan Snapp, Esq.

Pennsylvania Power

& Light Company 2 North Ninth Street Al 1 entewn, Pennsyl vani a 18101 Robert M. Gallo Resident Inspector P. 0.

Box 52 Shickshinny, Pennsylvania 18655 Susquehanna Environmental Advocates c/o Gerald Schultz, Esq.

500 South River Street Wilkes-Barre, PA 18702.

John L. Anderson Oak Ridge National Laboratory Union Carbide Corporation Bldg. 3500, P. 0.

Box X

Oak Ridge, Tennessee 37830 Nr. E. B. Poser Project Engineer'echtelPower'orporation PE 0.'Box 3965 San:Francisco, California 94119 t1atias F. Travieso-Diaz,

.Esq.

Shaw, Pittman, Potts Trowbridge 1800 M St reet, N.

W.

Washington, D. C.

20036 Dr. Judith H. Johnsrud Co-Director Environmental Coalition on Nuclear Power 433 Orlando Avenue State College, PA 16801 Mr. Thomas M. Gerusky, Director Bureau of Radiation Protection Department of Environmental Resources Commonwealth of Pennsylvania P. 0.

Box 2063 Harrisburg, PA 17120 Ms. Colleen Marsh Box 538A, RD¹4 Mountain Top, PA 18707 Mrs. Irene Lemanowicz, Chairperson The Citizens Agai'nst Nuclear Dangers P. 0.

Box 377 RD¹1

Berwick, PA 18503 Mr. J.

W. Millard Project Manager Mail Code 394 General Electric Company 1?5 Curtner Avenue San Jose, California 95125

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 sEPTEMBER I 9 I980 TO:

SUBJECT:

All Licensees of Operating Reactor Plants and Applicants for Operating Licenses and Holders of Construction Permits ADDENDUM TO THE CLARIFICATION LETTER FOR TMI ACTION PLAN REQUIREMENTS By letter dated September 5, 1980, we transmitted a preliminary clarifi-cation of the TMI Action Plan requirements.

Attached is a set of errata sheets which amend the referenced letter (viz., missing pages, scheduling, Tech Spec considerations, etc.).

Also included is a corrected table of the Implementation schedule.

It is our intention to develop and issue model technical specifications after issuance of the final requirements package.

incerely,

Enclosure:

As stated isen u

irector Division of Licensing Office of Nuclear Reactor Regulation cc w/enclosure:

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d.

An evaluation, including proposed

actions, on the conformance of the inadequate core cooling instrumentation system to Regulatory Guide 1.97, Rev.

2.

Any deviations should be justified.

e.

A description of the computer functions associated with ICC monitoring and functional specifications for relevant software in the process computer and other pertinent calculators, The reliability of nonredundant comp'uters used in the system should be addressed.

f.

A current schedule, including contingencies, for installation, testing and calibration, and implementation of any proposed new instrumentation or information displays.

g.

Guidelines for use of the additional instrumentation, and analyses used to develop these procedures.

h.

A summary of key operator action instructions in the current emergency procedures for inadequate core cooling and a description of how these procedures will be modified when the final monitoring system is implemented.

A description and schedule commitment for any additional submittals which are needed to support the acceptability of the proposed final instrumentation system and emergency procedures for inadequate core cooling.

TECHNICAL SPECIFICATION CHANGES RE UIRED Yes.

REFERENCES 1.

NUREG-0578 (Recommendation

2. 1.3.b).

2.

H. Denton (NRC) letter to All Operating Nuclear Power Plants on "Discussion of Lessons Learned Short Term Requirements,"

dated October 30, 1979.

EMERGENCY POWER FOR PRESSURIZER E UIPMENT (II.G. 1 POSITION Consistent with satisfying the requirements of General Design Criteria 10, 14, 15, 17 and 20 of Appendix A to 10 CFR Part 50 for the event of loss-of-offsite power, the following positions shall be implemented:

1.

Motive and control components of the power-operated relief valves (PORVs) shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.

2.

Motive and control components associated with the PORV block valves shall be capable of being supplied from either the offsite power source or the emergency power source when the offsite power is not available.

3.

Motive and control power connections to the emergency buses for th'e PORVs and their associated block valves shall be through devices that have been qualified in accordance with safety-grade requirements.

4.

The pressurizer level indication instrument channels shall be powered from the vital instrument buses.

The buses shall have the capability of being supplied from either the offs1te power source or the emergency power source when offsite power is not available.

CLARIFICATION 1.

While the prevalent-consideration from TMI Lessons Learned is being able to close the PORV/block valves, the design should retain, to the extent practical, the capability to open these valves.

2.

The motive and control power for the block valve should be supplied from an emergency power bus different from that which supplies the PORV.

3.

Any changeover of the PORV and block valve motive and control power from the normal offsite power to the emergency onsite power is to be accomplished manually in the control room.

4.

For those designs where instrument air is needed for operation, the electrical power supply requirements should be capable of being manually connected to the emergency power sources.

APPLICABILITY All PWR Operating License Applicants

-2" IMPLEMENTATION Prior.-to.,the:-issuance..of a fuel load license.

DOCUMENTATION RE UIRED Each applicant.,shall provide, sufficient documentation to support a reasonable assurance finding,~by the.,NRC,:,that, each of the posi.tions stated

'above are met; The.~documentation, should.include as a minimum,;supporting information including system:.design description,; logic;diagrams, electrical schematics, test.proce-duj;es;. and techni,cal;.: speci.fications; REFERENCES'UREG".0578,.

(Recommendation,:2. 1..1)

NUREG-.0694

- (Parti',1)

.'UREG-0660, (Section"II'.G. 1).',

CONTROL OF AFW INDEPENDENT OF ICS II ~ K. 2 ~ 2)

POSITION For 88W-designed

reactors, provide procedures and training to initiate and control auxiliary feedwater independent of the integrated control system (ISC)

CLARIFICATION None required APPLICABILITY All Operating License Applicants with 88W-designed reactors IMPLEMENTATION Prior to issuance of a full power license DOCUMENTATION RE UIRED Applicants shall provide sufficient documentation at leat four months prior to the issuance of a full power license to support a reasonable assurance finding by the NRC that the position specified above has been met TECHNICAL SPECIFICATIONS RE UIRED No.

REFERENCES NUREG-0660, (Section II.K. 2, Table C. 2, Item 2)

NUREG-0694, (Part II)

AUXILIARYFEEDWATER SYSTEM UPGRADING

~II. K. 2. 8)

POSITION All operating Babcock and Wilcox plants were ordered to be shut down shortly after the TMI-2 accident.

The Orders included both short-term and long-term actions.

The NRR Bulletins and Orders Task Force reviewed the licensees com-pliance with the short-term actions of the Orders and issued safety evaluation reports which served as the basis for plant restart.

Additional items were identified in the review of the long-term actions which requires further work by the licensees.

CLARIFICATION The licensees were required to comply with the Commission Orders regarding certain short-term and long-term AFWS modifications, The staff evaluated the short-term actions, and safety evaluations were prepared prior to the plants being allowed to return to operation.

The staff evaluation of the additional (long-term) items will be performed in conjunction with Item II.E. 1. 1, (Auxiliary Feedwater System Evaluation).

APPLICABILITY All B8W Operating Reactors.

IMPLEMENTATION No separate implementation is required for this item.

All AFW system upgrade modifications for B8W plants are being reviewed as part of Item II.E. l. l.

TYPE OF REVIEW See Item II.E.1.1.

DOCUMENTATION RE UIRED See Item II.E. l. l.

TECHNICAL SPECIFICATION CHANGES RE UIRED As required.

REFERENCES NUREG-0660, (Sections II.E. l. 1 and II.K. 2)

NUREG-0645, Volume 1, (Section 2.4.6)

~

f FAILURE MODE EFFECTS ANALYSIS ON ICS h

~(II. K. 2. 9 POSITION B8W licensees submit a failure mode and effects 'analysis (FMEA)'f the integrated control system (ICS).

,c CLARIFICATION A generic FMEA of the ICS (BAW-1564) was submitted on August 17, 1979 by the operating plant licensees.

This report was reviewed by the staff and ORNL.,

Requests for additional information, regarding the recommendations contained in the report, were sent to the licensees on November 7, 1979.

The responses to the November 7, 1979 letter have been received and are under review.

APPLICABILITY All 88W Operating Reactors and Operating License Applicants.

IMPLEMENTATION 0 eratin Reactors Open - Staff recommendations pending completion of staff review.

0 eratin License A

licants Prior to issuance of a full-power license.

TYPE OF REYIEW Postimp1 ementati on revi ew.

DOCUMENTATION RE UIRED 0 eratin Reactors To be determined fo11owing staff review.

0 eratin License A

licants B8W applicants should provide the following:

l.

Identify whether the previous generic submittal (BAW-1564) is applicable to your plant, and 2.

Specify what actions have been taken at your facility to comply with the recommendations listed in BAW-1564.

TECHNICAL SPECIFICATION CHANGES RE VIREO To be determined following staff review.

REFERENCES Commission Orders on B8W Plants "Integrated Control System Reliability Analysis," BAN-1564.

Letter from R; W. Heid (NRC) to All B8W Operating Plants, dated November 7, 1979.

" NUREG-0645, (Volume 1, Section 2.4.6)

'UREG-0694, (Part 2)

SAFETY-GRADE ANTICIPATORY REACTOR TRIP II. K. 2. 10)

POSITION Upgrade the currently installed control-grade, anticipatory reactor trip (ART) on loss-of-feedwater and turbine trip to safety-grade.

CLARIFICATION 0 eratin Reactors 1.

IE Bulletin 79-05B, Item 5, issued on April 21, 1979, directed B8W licensees to provide a design and schedule for implementation of a safety-grade reactor trip. upon:

a.

loss of feedwater; b,

turbine trip; and c.

significant reduction in steam generator level.

2.

In accordance with IE Bulletin 79-05B, the B8W licensees submitted a

conceptual design for a safety-grade, anticipatory reactor trip which would be initiated upon turbine trip and loss of feedwater only.

Included in the licensees'esponses was a generic evaluation prepared by B8W which proposed that the anticipatory reactor trip on low steam generator level was not necessary.

3.

Staff review of these submittals resulted in a preliminary design approval for the safety-grade anticipatory reactor trip being issued to the B8W licensees on December 20, 1979.

However, the approval letters also specified'he additional information which would be required to be sub-mitted prior to final staff approval of the design.

4.

The staff will complete its review of the generic evaluation by B8W which indicates that the proposed anticipatory trip on low steam generator level is unnecessary.

Further clarification will be provided on this matter, if required, following completion of the staff review.

0 eratin License A

licants Compliance with TMI Action Plan, Item II.K.1.21, satisfies this requirement.

APPLICABILITY All B8W Operating Reactors and Operating License Applicants.

!'iIMPLEMENTATION OATE

',10 eratin Reactors

."'Submission of fi,nal design information - October 1, 1980 liInstallation of safety"grade trip - June 30, 1981

'0 er atin License A

1icants

>>Implementation of TMI Action Plan Item II.K.1.21 prior to the issuance of

(.the fuel load satisfies this requirement.

ITYPE OF REVIEM

'Preimplementation Review.

'OOCUMENTATION RE UIRED "The'following information was identified as required by the staff for the

'final design approval, as noted in item 3 above:

1.

The final design.submittal should include the final logic diagrams, electrical schematic

diagrams, piping and ',instrumentation diagrams and location layout drawings.

'2;.

For sensors located in nonseismic areas which have not previously con-

,tained'RPS

inputs, perform and submit an analysis which shows that. the

',installation (including circuit routing) is.~designed such that the effects of credible faul.ts (i.e., grounding, shorting, application of high voltage, or electromagnetic interference) o'r failures in these areas could not be

'.propagated back,to the RPS and degrade the'RPS performance or operability.

'3.

Submit "Seismic and Environmental gualification Summary Reports" for the equipment which has not been previously submitted.

In addition, demonstrate

,that. the environmental test conditions bound the actual worst case accident conditions expected at.the installed locations.

-4.

Assure th'at the ARTs testability includes provisions to perform channel functional tests at power.

Testing of this circuitry is to be included in the RPS monthly surveillance tests.

.5.

< Include in the final design submittal the RPS check-out procedure which will demonstrate both the operability of the new trip circuitry and the continued operability of the previous RPS.

The above information should be submitted for staff review by October 1,

1980.

-3" TECHNICAL SPECIFICATION CHANGES RE UIRED Yes.

REFERENCES Commission Orders on 88W Plants IE Bulletion 79-058, Item 5 Letter from R.

W.

Reid (NRC) to 88W Licensees, date Oecember 2g, 1979

Subject:

Preliminary design approval for safety-grade anticipatory reactor trip and request for additional information.,

NUREG-0645, (Volume 1, Section 2.4.6)

NUREG-0694, (Part 2)

CONTINUED OPERATOR TRAINING AND DRILLING

'I.K.2.11 POSITION

~

Continue operator training and drilling to assure a high state of preparedness.

C LARIFICATION In a letter from D.

F.

Ross, Jr.

(NRC) to All B8W Operating Plants, dated August 21, 19?9, each B8M licensee was requested to document the steps they had taken to insure that continued operator training and drilling incorporated the necessary lessons learned from the accident at THI-2.

This response was required to assure compliance with the long-term training requirements of the Commission Orders.

Responses to this request were received from the licensees and reviewed by the NRC staff.

Based on that review, the staff concluded that the training programs had been sufficiently modified to incorporate the necessary lessons learned from THI such that this portion of the Commission Orders was satisfied.

A.

complete evaluation of this item is discussed in Section 2.4.6 of NUREG-0645, Volume l.

Additional requirements, beyond the intent of the Commission Orders, are being implemented through the following items of the Action Plan: I.A.2.2, I.A.2.5, I.A.3.1, and I.G.l.

APPLICABILITY Al 1 B8M Operating Reactors IMPLEMENTATION DATE COMPLETED TYPE OF REVIEW Postimplementation review.

DOCUMENTATION RE UIRED No additional documentation required.

TECHNICAL SPECIFICATIONS RE UIRED No.

'PREFERENCES

-L'etter"from,.D.

F::. Ross Jr., (NRC),,:to,,ALL BABCOCK-:8.WILCOX OPERATING PLANTS

. l(EXCEPT".THREE HZL'E.:ISL'ANO, UNITS,1 &,2),.dated August 21, 1979,

Subject:

..~Identi'ication -and,;Resol'ution; of, Long;Term,'eneric Issues Related to the

.:.Commission, Orders-..of,May,1979.

l~NUREG'=0645, !lReporti;of. the; Biil,letins & Orders'ask Force,"

Vol,ume I, January 1980.

~i:,PUREG:0660, Section,'II; K; 2,;".Item.C;.11.

THERMAL MECHANICAL REPORT-EFFECT OF HPI ON VESSEL INTEGRITY FOR SMALL BREAK LOCA WITH NO AFW II. K. 2. 13 POSITION Perform a detailed analysis of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater.

CLARIFICATION The position deals. with the potential for thermal shock of 88W reactor vessels resulting from cold safety injection flow.

One aspect that bears heavily on the effects of safety injection flow is the mixing of safety injection water with reactor coolant in the reactor vessel.

B&W has committed to provide a

report by the end of July which will discuss the mixing question and the basis for a conservative analysis of the potential for. thermal shock to the reactor vessel.

APPLICABILITY All B8W Operating Reactors and Operating License Applicants.

IMPLEMENTATION DATE Confirmatory information requested.

Implementation of any modifications will be subject to the results of NRC staff review of the report.

TYPE OF REVIEW Postimplementation Review.

DOCUMENTATION RE UIRED Licensees shall submit results of evaluation by January 1,

1981.

Applicants shall submit results of evaluation at least four months prior to the issuance of a full power license.

TECHNICAL SPECIFICATION CHANGES RE UIRED To be determined following staff review.

REFERENCE NUREG-0645, (Volume 1 Section 2.4.5)

Letter from D.

F.

Ross Jr.

(NRC) to All B8W Operating Plants, dated August 21, 1979.

EFFECTS OF SLUG FLOW ON STEAM GENERATOR TUBES II.K.2.15 POSITION While the staff believed that the potential for slug flow was not great in 8&W

plants, because of the venting path provided by the internal vent val ves, the staff required a confirmatory evaluation of the effects of slug flow on steam generator tubes be performed by the licensees to assure that the tubes could withstand any mechanical loading which could result from slug flow.

CLARIFICATION V

The request for this information was originally sent to the B8W licensees in a letter from R.

W. Reid (NRC) to All B8W Operating Plants dated November 21, 1979.

The results of this analysis has been submitted by the licensees and is presently undergoing NRC staff review.

APPLICABILITY All B8W Operating Reactors and Operating License Applicants.

IMPLEMENTATION DATE Confirmatory information requested.

Implementation of any modifications will be subject to the results of NRC staff review of the evaluation.

TYPE OF REVIEW Postimplementation.

DOCUMENTATION RE UIRED No additional documentation is required at this time from Licensees.

Applicants must supply the requested.information at least four months prior to issuance of a full power license.

TECHNICAL SPECIFICATION CHANGES RE UIRED No.

REFERENCES Letter from R.

W.

Reid (NRC) to All 88W Operating Plants, dated November 21, 1979.

NUREG-0565, (Recommendation 2.6.2. 1)

NUREG-0645, (Volume 1, Section 2.4. 6)

NUREG-0694, (Part 2)

REACTOR COOLANT PUMP SEAL DAMAGE II. K. 2. 16 POSITION Evaluate the impact of reactor coolant pump seal damage and leakage due to loss of seal cooling upon loss of offsite power.

If damage cannot be precluded, licensees should provide an analysis of the limiting small-break LOCA with subsequent RCP seal damage.

CLARIFICATION The request for this information was-originally sent to the B8W licensees in a letter from R.

W.

Reid (NRC) to All B8W Operating Plants dated November 21, 1979.

The results of these evaluations have been submitted by the licensees and are presently undergoing NRC staff review.

'APP LICABILITY All B8W Operating Reactors and Operating License Applicants.

IMPLEMENTATION DATE Confirmatory information requested.

Implementation of any modifications wi 11 be subject to the results of NRC staff review of the evaluations.

I TYPE OF REVIEW Postimplementation.

DOCUMENTATION RE UIRED No additional documentation is required at this time from Licensees.

Applicants shall submit.the.requested information at least four months prior to the issuance of a full power license.

TECHNICAL SPECIFICATION CHANGES RE UIRED No.

REFERENCES Letter from R.

W.

Reid (NRC) to All B8W Operating Plants, dated November 21, 1979.

~ NUREG-0565, (Recommendation 2.6.2.f)

NUREG-0645, (Volume 1, Section 2.4.6)

NUREG-0694 (Part 2)

POTENTIAL FOR VOIDING IN THE RCS DURING TRANSIENTS II.K.2.17 POSITION Analyze the potential for voiding in the reactor coolant system during antici-pated transients.

CLARIFICATION The background for this concern and a request for this analysis was originally sent to the B&W licensees in a letter from R.. W.

Reid (NRC) to All 8&W Operating

Plants, dated January 9,

1980.

The results of this evaluation has been submitted by the 8&W licensees and is presently undergoing staff review.

APPLICIABILITY All 8&W Operating Reactors.

IMPLEMENTATION DATE Confirmatory information requested.

Implementation of any modifications will be subject to the results of NRC staff review of the licensees'valuation.

TYPE OF REVIEW Postimplementation Review.

DOCUMENTATION. RE UIRED No additional documentation is required at this time.

TECHNICAL SPECIFICATION CHANGES RE UIRED REFERENCE Letter from R.

W.

Reid (NRC) to All 8&W Operating Plants, dated January 9,

1980.

NUREG-0660,Section II.K.2 Item C. 17.

SE UENTIAL AFW FLOW ANALYSIS II.K. 2.19)

POSITION Provide a benchmark analysis of sequential auxiliary feedwater flow to the steam generators following a loss of main feedwater.

CLARIFICATION This requirement was originally sent to the B&W licensees in a letter from D.

F.

Ross Jr.

(NRC) to All B&W Operating Plants, dated August 21, 1979.

The results of this analysis has been submitted by the B&W licensees and is presently undergoing staff review.

APPLICABILITY All B8W Operating Reactors.

IMPLEMENTATION DATE Confirmatory information requested.

Implementation of any modifications will be subject to the results of NRC staff review of this analysis.

TYPE OF REVIEW P os timp 1 ementati on Rev iew.

DOCUMENTATION RE UIRED No additional documentation is required at this time.

TECHNICAL SPECIFICATION CHANGES RE VIREO No.

REFERENCE Letter from D.

F.

Ross Jr.

(NRC) to All 88W Operating Plants, dated August 21, 1979.

NUREG-0645, Volume 1, Section 2.4.6.

SMALL-BREAK LOCA WHICH REPRESSURIZES THE RCS TO THE PORV SE'TPOINT

( II. K. 2. 20)

'POSITION Provide an.analysis which shows the plant response to a small break loss-of-coolant accident.during which the reactor coolant system is repressurized to the ~PORV setpoint with subsequent failure of the PORV to close.

,'C L'A'RIF.I CATION

'The requirements.was originally sent to the 88W licensees in a letter from

D..

F.:Ross Jr.

(NRC) to All 88W,Operating Plants, dated August 21, 1979.

'The results of this "analysis has been submitted by the BKW licensees and is

presen'tly.undergoing staff review.

'APPLICABILITY

'A11.88W Operating Reactor s.

ZMPLEME~TATION DATE

Confirmatory informat'ion requested.

Implementation of any modifications will

be the subject to the results of NRC staff evaluation of this analysis.'YPE

'OF. REVIEW

'Postimplementation Review.

'DOCUMENTATION RE UIRED

'No additional documentation is required at this time.

TECHNI'CAL SPECIFICATION CHANGES RE VIREO No.

REFERENCES Letter from D.

F'.

Ross Jr.

(NRC) to All 88W Operating Plants, dated August 21, 1979.

NUREG-0565, Recommendation 2.6.2.c NUREG-0645,

.Vol.ume 1,.Section 2.4.6