ML18017B261

Part 21 Rept Re Flexing of Unit 1 Containment Dome Due to Imposition of Uneven Loads During Movement for Final Placement.Evaluation Re Reportability Per 10CFR50.55e Not Completed.Evaluation to Be Completed by 801115

ML18017B261
Person / Time
Site:	Harris
Issue date:	09/23/1980
From:	Chiangi N CAROLINA POWER & LIGHT CO.
To:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
REF-PT21-80-288-000 10CFR-050.55E, 10CFR-50.55E, SH-N-2-18, NUDOCS 8010230185
Download: ML18017B261 (24)

Text

~4llfV-PoLQLv p+T 2I IIp(TIFICATIp),

%Xi-z.~-toccl ClR W"E DATE OF LETIIg 4~ %so IKRtH NO.

oo-<co <o'oduo5 IIATE.DISTRIBUTED (o4z. m

~.

ORISINAL REPORT)L SUPPLEI'ENTARY DISTRIBUTION:

REACTOR (IO

)C F(H. CYCLE S l%TERIALS (N)

IE FIGS AQ/FR'EI RZGIOXIS I,II,III,IY,Y YENDOR BR, R-IY t'KS / FCNS SS-5F)

LOG / (PA N'ta 5715 ABE NS 7602 ASLBI'. E/H 050 SAKCIIARDS (S)

IE FILES AQ/SG AQ/ROI REGIONS I,II,III,IY,Y VHSOR BR. R-IY NRR/KL K / SG SS-881 LDB / (PA t"tS 5715 AEOD NS 7602 IE FILES AQ/ROI AQ/RCI RHIOt6 I,II,III,IY,Y VQKGR BR, R-IY LOS / N'A llS 5715 AEOD NB 782 NRR/KE NRR/OSI rIR/DST NRR/KL SAP/SP.

t'%-7210A CORAL FILES 016 CENTRAL FIL~S ( CHRON)

PDR PDR TERA CENTRAL FILES 016 CEtRRAL FILES (CHROI CENTRAL FILES SS->96 PDR LPDR TERA CB'TRAL FILES 016

, m TRAL>>LES (CHtnz PDR LPDR TEII ACTION:

t4'SS

~

~

CP PRELIMINARY EVALUATION OF THE AITACHED REPORT INDICATES LEAD RESPONSIBILITY FOR FOLLOHJP AS SMWI BELOW:

FFMS I 80 l(I23eiFfi PZi, 8/1/80 z/iqfso

gp,S RE0y (4

0 l

.~)( I 0ji e

+a*++

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SEPTEhlBER 1 9 ]980 g(p 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:

C I

1 a

Py

> ~

OPERATING REAC'IOR AEIRIIRBlENTS INPLENEIITATION SCNEOIILE (As of Septeeber 5, 1980) l.h.l.3(V)

I.A.2. 1(V)

I.A.3.1(P)

II.K.3(P) l.h. l. l{l) i.h.i.2 l.h. JP. I(I')

i.h-.i. S l.h; S.i(V)

I.C. I ( l')

I.C.P I

I

'e I

I l l l.I:.:(I')

lI.I:. L(V) li.u.l(V) l i.i2.'~(l )

Il.l.I.2(V)-

Il.l.l. l(l')

II.I.. l. l(l')

I i.1..4.2(V)

Il.I:.2( V)

II.G.I(l')

l I.K. 1 (I')

I I.K.2(V) ill.h.l. l(P) lll.h.l.2(l')

III.D.I.I(V)

Ill.0.3.S(P) l TT.VI o2.82 h.

EPI 'I/Uj lk.L-I)2:2. l9fS

~lil22-Hdl 1981 I.A.I.I(P)

I.c.l(P)

I.C.5 L.cu6 11.0.3(P)

II.E.1.2(V)

II.E.3.1(P)

II.E.4.2(P)

II.F. 1(P)

II.F.2(P)

II.G.I(P)

I I.K. I( P)

II.K.2( P)

II.K.3(P) 111.A.1.2(P)

L I1.0.1.1(P) 111.0.3.3(P) 111.0.3.4(V)

II.B.4 II.B.I(P)

II.E.4.1(P) 11.0.1(P)

II.K.2(P)

II.E.l.l(P)

II.K.3(P)

IL.E.4.2(P)

III.A.1.1(P)

II.K.3(P)

'.C.1(P) 11.F. 1(P)

ANr-June 1981 J~ut-Se t 1981 Oct-Oec 1981 Jan.Nar 1982

~Ar 1982 Jul 1982 II.B.I (P)

IIL.A.1.2(P)

I.A.1.3(P)

II.S.2(P)

II.D.1(P)

II.B.3(P)

II.E.1.1(P)

LI.F.2(P)

LI;K.3(P)

Jan 1983 II.K.3(P)

III.D.3.4(P)

I.A. I. I(V)

I.A. I.v I.A. I. 5 I.n.'5. l ( v)

I.ls.1.2 I.I. 1(l']

I.I.. n'

.I.. '5

. I.L.4 I.I../(I')

I.D. I (I')

I.r.. I(I')

I I. 15.4(I')

I I.0.1( V)

Il.0.3 II.I. I.2(I')

I I. I.4. I(I')

II.I. I(l')

II.I.2(I )

) I.li. I I I. V. I(I')

II.f. 5(l >

I I I.A. I. I ( I')

III.A.I.V(l)

III.D.5. 5! V)

I o I I I (a((.r I.L.7(l )

I.c.0(v)

I.G. I (I')

I I. II. I ( I')

11.15.2(V) 11.0.3(l )

11.0.4(V)

I I.I.. I. I II.E.3.1 II.L.4.2 II.K.2(P)

II.K.3.3 lll.h.1.1(P)

III.D.I.I lll.0.3.4 0(.t-l)e<. I900 I I.U.IA I I.K.3(P)

JJD)-Har 1981 I.A.I. I (P)

I.C. 1(P)

'I.E.I.2(P)

II.E.4.2(P)

>>.F.1(P)

I I.K.3(P) lll.h.1.2(P) lll.D.3.3(P)

OL REQU IREHEIITS As of Septa))ber 1980 Apr-June 1981

~Jul

-Se t 1981 II.E.4.1(P)

I I.D.1(P)

II.K.3 II.E.4.2(l)

II.K.3(P)

Ont-nee 1981 Jan-Mar 1982

~A r-June 1981 Jen 1.

1983 I.C.I(P) 11.8.1(P)

III.A.1.2(P)

II.K.S(P)

II.F.1(P)

I I.B.2(P)

III.D.3.4(P)

I I.B.3(P)

II.D.1(P)

II.F.2(P)

II.K.3(P)

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 computers 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 Oesign Criteria 10, 14, 15, 17 and 20 of Appendix A to 10 CFR Part 50 for the ev:nt of loss-of-offsite

power, the following positions shall be implemented:

1.

Motive and control components of the power-operaied 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 the 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 offsite 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 requi rements should be capable of being manually connected to the emergency power sources.

APPLICABILITY All PWR Operating License Applicants

"2-I

'IMPLEMENTATION Prior to the issuance of a fuel load license.

DOCUMENTATION RE VIREO Each applicant shall provide sufficient documentation to support a reasonable assurance findirig by the NRC that each of the positions stated above are-met.

T)e documentation should include as a minimum, supporting information including system design description, logic. diagrams, electrical schematics, test proce"

dures, and technical specifications.

REFERENCES IIIIREG-0578, (Rec'ommendation

2. 1. 1)

N0REG"0694, (Part 1)

NUREG-0660, (Section II.G. 1)

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

POSITION For B8W-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 B8W-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 B&W Operating Reactors.

IMPLEMENTATION No separate implementation is required for this item.

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

TYPE OF REVIEW See Item II.E ~ 1.1.

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

TECHNICAL SPECIFICATION CHANGES RE VIREO As required.

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

NUREG-0645, Volume 1, (Section 2.4.6)

i FAILURE MODE EFFECTS ANALYSIS ON ICS

~II. K. 2. 9 POSITION BKW licensees submit a fdilure mode-and effects analysis (FLEA)'f the. integrated control system (ICS).

CLARIFICATION i

A generic FMEA of the ICS (BAM-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 BEM Operating Reactors and Operating License Applicants.

IMPLEHENTATION 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 REVIEW Postimplementation review.

DOCUMENTATION RE VIREO 0 eratin Reactors To be determined fo1lowing staff review.

0 eratin License A

licants B8W applicants should provide the following:

1.

Identify whether the previous generic submittal (BAM-1S64) is applicable to your plant, and 2.

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

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

REFERENCES Commission Orders on B8M Plants "Integrated Control System Reliability Analysis," BAN-1564.

Letter from R.

M. Reid (NRC) to All BN Operating Plants, dated November 7, 1979.

NUREG-0645, (Volume 1, Section 2.4.6)

NUREG-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 l.

IE Bulletin 79-05B, Item 5, issued on April 21, 1979, di rected B8M 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 B8M 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 B&W licensees on December 20, 1979.

However, the approval letters also specified'the 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 B8M 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.

"2" IMPLEMENTATION DATE 0 eratin Reactors Submission of final design information " October l, 1980 I'nstallation of safety-grade trip - June 30, 1981 0 eratin License A

licants Implementation of TMI Action Plan Item II.K. 1. 21 prior to the issuance of the fuel load satisfies this requirement.

TYPE-OF REVIEW Preimplementation Review.

DOCUMENTATION RE UIRED The fol:lowing 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 faults (i.e., grounding, shorting, application of high voltage, or electromagnetic interference) or 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'eports" 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 that 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 ci rcuitry and the continued operability of the previous RPS.

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

1980.

TECHNICnl SPECIFICATION CHANGES RE UIRED Yes.

REFERENCES Commission Orders on B8W Plants IE Bulletion 79-058, Item 5 Letter from R.

W.

Reid (NRC) to B8W 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 II.K.2.11 POSITION Continue operator training and drilling to assure a high state of preparedness.

CLARIFICATION In a letter from 0.

F.

Ross, Ji.

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

1979, each 88W 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 TMI-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 TMI 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 1.

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.1.

APPLICABILITY All 88W Operating Reactors IMPLEMENTATION DATE COMPLETED TYPE OF REVIEW Postimplementation

review, DOCUMENTATION RE UIRED No additional documentation required.

TECHNICAL SPECIFICATIONS RE UIRED No.

INFERENCES L<e'tter from D.

F.

Ross Jr.

(NRC), to ALL BABCOCK 8 WILCOX OPERATING PLANTS

(~EXCEPT THREE NIL'E'ISLAND, UNITS 1 &: 2), dated August 21, 1979,

Subject:

Identification'h'd Resolution of Long-Term Generic', Issues Related to the dommission Order's of May 1979.

hllLIREG"'0645, "Rep'ort of the Bulletins 8 Orders Task Force,"

Volume I, January 1980.

HlJREG-0660,Section II.K.2, Item C.ll.

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 B8W 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.

88W 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 B8W

plants, because of the venting path provided by the internal vent valves, 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 The request for this information was originally sent to the B8W licensees in a letter from R.

W. Reid (NRC) to All 88W 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 88W 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 VIREO 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 VIREO No.

REFERENCES Letter from R.

W. Reid (NRC) to All B8W 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)

(

POSITION REACTOR COOLANT PUMP SEAL DAMAGE II. K. 2. 16 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 ~i 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.

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 evaluations.

TYPE OF REVIEW Postimplementation.

DOCUMENTATION RE UIRED No additional documentationis 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 3., Section 2.4.6)

NUREG-0694 (Part 2)

I 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 B8M licensees in a letter from R., M. Reid (NRC) to All BEW Operating

Plants, dated January 9,

1980.

The results of this evaluation has been submitted by the B8M licensees and is presently undergoing staff review.

APPLI C IABILITY All 88W 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 No.

REFERENCE Letter from R.

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

1980.

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

ft I

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 B8W 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 88W 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 Postimplementation Review.

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

TECHNICAL SPECIFICATION CHANGES RE VIREO 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 SETPOINT

( II. K. 2. 20)

, POSITION Provi'de "an anal'ysis which shows the plant response to a small break loss-of-

'oolant accident during which'he reactor coolant system is repressurized to the PORV setpoint with subse'quent failure of the PORV to close.

.'CLARIFICATION The~'requireme'nts was originally sent to the 88W licensees in a letter from

.D.-

F.

Ross Jr.

(NRC).to A11.88W Operating Plants, dated August 21, 1979.

The'".re'suits."of this analysis has been submitted by the 88W licensees and is presently. undergoing staff review.

APPLICABILITY

'l 1

. 88W Oper ating~ Reactors.

... IMPLEME "TATION".DATE

- Confirmatory information requested.

Implementation of any modifications will

'be'the s'ubject-to. the-results of NRC staff evaluation of this analysis.

,, TYP.E'-'OF REVIEW "Postimplementation Review.

, 'OOCUMENTATION RE UIRED

'No-'a'dditional 'documentation is required at this time.

-TECHNICAL SPECIFICATION CHANGES RE UIRED

'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, Volume 1, Section 2.4.6