ML20003A171

Safety Evaluation Supporting Amend 63 to License DPR-65

ML20003A171
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Site:	Millstone
Issue date:	01/14/1981
From:	Office of Nuclear Reactor Regulation
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NUDOCS 8101290804
Download: ML20003A171 (14)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT h0. 63 TO FACILITY OPERATING LICENSE NO. OP NORTHEAST NUCLEAR ENERGY COMPANY, ET AL.

MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 DOCKET NO. 50-336 Introduction Early in the review of the Three Mile Island Unit No. 2 (TMI-2) accident, it became apparent that increased plant safety would result frem automatic This was short-tenn initiation of auxiliary feedwater system (AFWS) flow.

recommendation No. 2.1.7a of our July 1979 NUREG-0578.

In the implementation 30, 1979, we provided clari'ication letters dated September 13 and October of requirement No. 2.1.7a and proposed control grade system installation by January 1,1980 with the upgrading of the automatic initiation of AFWS flow to safety grade by January 1,1981.

Northeast Nuclear Energy Company In letters dated November 21 and 30,1979, (NNECO) pointed out that modifying the AFWS to be automatically initiated was not necessary and constituted an unreviewed safety question issue since AFWS flow was not considered in the Millstone Unit No. 2 (M111 stone-2) main steam-NNECO (and other licensees) contend that the

.line break (MSLB) analyses.

addition of AFW flow during a MSLB accident will: (1) result in a positive reactivity insertion (due to increased cooldown) and, thus, a higher final return-to-power condition; and (2) a higher peak containment pressure than the values calculated in the analysis of record.

Reiterating their concern about the unreviewed safety question, NNECO proposed, by letters dated their control design for autcmatic initiation of December 6 and 17,1979, AFWS flow.

address the NNECO concern. We agreed Our letters of December 21 and 27,1979 that AFWS flow may adversely affect the MSLB accident and requested a re-analysis of this accident to be submitted for our review prior to the finalThe connection of the circuits involved to automatically initiate AFWS flow.

25, 1980 as requested reanalysis was supplied by the NNECO letter of JanuaryThis Safety Eval 11, 1980.

supplemented by letter of Aprilreview the effects of automatic initiation of AF of return to power (SE Section 2.1) and on the calculated peak containment The pressure (SE Section 2.2) during the main steamline break a and accidents will be addressed in SE Section 2.3.

In the application for Technical Specifications (TS) changes for automatic initiation of the AFWS, NNECO provided an analysis redefining the pump rated The capacity. This analysis will be evaluated in Section 2.4 of this SE.

proposed TS changes for the AFWS applications dated March 31, May 20 and August 29, 1980 will be addressed in Section 2.6.

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} Our letter of October 22, 1979 presented the staff reliability evaluation of NNECO's the M111 stone-2 AFWS and made short-term and long-term recommendations.

responses to these recommendations were submitted by letters dated November 28, 1979 and January 17, March 10, and June 16, 1980. Our review of tnis infomation is presented in Section 2.5 of this SE.

The detailed review of the safety grade instrumentation system required to automatically initiate the AFWS is to be reviewed by the Franklin Research Center in Philadelphia, Pennsylvania, under NRC contract. The resultant Safety Evaluation will be issued at a later time.

2.0 Discussion and Evaluation 2.1 "SLB Accident - Return to Pcwer NNECO's analysis of the effects of return to power following a MSL3 accident is presented in Attachment 1 to their January 25, 1980 letter. The starting conservative assumptions, according to NNECO, for this analysis are:

Only a three-minute delay in delivery of auxiliary feedwater flow to the e

steam generators was assumed, rather than a more realistic longer time del ay,

Credit is not taken for complete isolation of the main feedwater system, e

thereby resulting in a continuous flow of 772 gpm (5 percent of full flow) of main feedwater to the affected steam generator, A conservative representation of auxiliary pump feedwater flow, namely e

2800 gpm, which is 35% higher than maximum runout flow at Millstone Unit No. 2.

Thus, a total of 3572 gpm of feedwater flow is assumed in the

analysis, e Failure of one HPSI pump, t

e Failure of one LPSI pump, The highest worth CEA is assumed to stick in the fully withdrawn position, e

j and j

The end of Cycle 3 moderator temperature and Doppler (fuel temperature) s coefficient values were used since these values result in the greatest positive reactivity change during cooldown.

The analysis assumed that the event is initiated by a circumferential rupture of a 34 inch main steam line at the steam generator nozzle. NNECO states that this break is limiting since it results in the greatest rate of temperature reduction in the reactor core region. The reanalysis reported jn the January 25, 1960 submittal uses the same assumptions and methods as previously used except that it simulates automatic initiation of auxiliary feedwater flow in three minutes from initiation of the event.

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i The rationale for delaying the initiation of AF45 originates from the positive reactivity feedback which accompanies a postulated MSLE. During a pcstulated double-ended guillotine break of this steam line, the br0 ken steam generator behaves as an enhanced heat sink, resulting in rapid cooldown of the primary This rapid cooldewn has a noticeable impact on 2e moderator reactivity system A conservative feedback, wnich results in a net positive reactivity insertien.

assu ption is made that te limiting control element asse-bly (CEA) is stuck in its fully withdrai<n ecsition.

Based on the licensee's generic analyses, the reactivity feecback was mest limiting For this conditien, the calculations predict that during full pcwer operaticn.

there would be a 10.St return to pcwer due to the ecoling effect of the auxiliary feedwater; hewever, this is less tnan the 12 return to power precicted ;rior-to-auxiliary feeowater injection. The net energy reecved frcct the krimary system was conservatively assumed to be the product of the total steam gestrator secondary Sheuld lipuid entrainment mass (NOT) times the latent heat of evaporation (nfg).

exit the break, then the energy removed from the primary system will be less severe.

For a postulated guillotine break in a steam line, the time recuired to deplete ce breken steam generator secondary inventory is accroximately 1 1/2 minutes (for Should the auxiliary feedwater inject into the steam

.the full power condition).

generator immediately, when called upon to do so, then the magnitude of the primary side cooldown is increased ('4707 x nfg; where hot is increased). This results in enhancing the primary side cooling and in an increased reactivity The mechanism available for turning the reactivity around is the feedback.

initiation of ECCS, wnich injects boron into the system.

Due to the time constraints in providing analytical assessment of auto-initiation This af ne AF45, a generic review was conducted by Combustion Engineering (CE).

that a three-minute celay in the initiation of AFWS will ensure review cencluded that the ONSR limit will not be exceeded. Fuel rods =nien exceed tne DNSR limi are assu: red to fail (a conservative assumotion).

l The purpose of the tnree-minute delay is to provide time for the ECCS injected l

berated water to itssen the magnitude of the moderator reactivity 'eedback I

4-attributed to the AFWS inventory. Analyses have shown that during full power operation, the core becomes critical during the blowdown of a steam generator.

Af ter the steam generator has blown dry, the auxiliary feedwater injects, thus creating a second return to criticality, but at a magnitude less tnan experienced during the blowdown phase.

The licensee's analytical method for analyzing steam line breaks is presently under The review at this time indicates resaonable assurance that the staff review.

conclusions based on the submitted analyses will not be appreciably altered by The staff finds the return to the completion of the analytical methods review.

power results following a MSLB accident with automatically initiated AFWS flow delayed three minutes are not more limiting than previous analysis results without automatic AFWS flow and are, therefore, acceptable.

NNECL states that single f ailures concurrent with the MSLB, other than those If st.ed in the assumptions, as well as loss of offsite power concurrent with MSL3, are not and have not been part of the design basis as described in the FSAR and, therefore, were not considered.

Other significant failure would be one that results in a higher fbedwater flow that was assumed in the present analysis. The present analysis assumed that main feedwater flow was reduced from 100 to 5 percent in 60 seconds by the In reality, the feedwater flow would be reduced feedwater control system.

from 100 to O percent in less than 10 seconds by the action of two safety These systems, the main feedwater isolation and the main grade systems.

steam isolation valves, are actuated by low steam generator pressure and cause the closure of the main feedwater isolation valves and main feedwater pump By not taking credit for these systems, the present analysis contains trip.

From our review, we conclude that single a multiple failure assumption.

failure has been treated in an acceptable manner in this analysis for Millstone 2.

The primary consequences resulting from loss-of-offsite power (LCOP) are a delay of emergency core cooling systems (ECCS) injection and tripping of During LOOP, ECCS injection is delayed approxi-the reactor coolant pumps.

mately 25 seconds as the emergency diesel generators restore power to the LOOP also results in coastdown of the reactor coolant pumps.

ECCS pumps.

Continued operation of the reactor coolant pumps would have two effects on an SLB transient:

Running the reactor coolant pumps (RCPs) results in a greater degree of overcooling as the hot primary fluid is forced through the steam e

generators, and The reactor coolant pumps act as a driving head, forcing the ECCS a

injected borated water into the core.

Thus, losing offsite pcwer affects the degree of system cooling and the time at wnich the ECCS-injected bcron enters the reactor core. Overcooling and berated water injection are competing effects in which the forner increases reactivity and the latter reducc: reactivity.

In reviewing past analyses of MRB for other plants similar to Millstone-2, we have determined that reduction in the RCS cocidewn rate caused by coastdown of the RCP after LOO' has had a larger effect than slower boren injection to the core.

Thus, we find that the MSLB accident is reduced in severity with a concurrent loss of offsite power.

We find automatic initiation of the auxiliary feedwater system to inject needed makeup water to the steam generators withcut the need for operator action will improve the nuclear safety of Millstone-2. The staff plans to perform independent audit calculations by the end of FY 81 to provide further confirmation of our conclusions.

2.2 MSL3 Accident - Peak Containment Pressure f

Appendix 5 of NNECO's January 25, 1980 letter provides a response to questions posed by our letter of December 21, 1979.

Specifically, NNEC0 was to assess the potential for containment overpressurization due to the anticipated enntinuous addition, at pump runcut flow, of auxiliary feedwater to the affected steam generator following a postulated MSLB accident.

Automating the auxiliary feedwater system would cause an increase in energy released to containment af ter a MSLB thereby increasing the containment pressure.

The original FSAR analysis of the MSLB accident was based on the no load, single loop nc::el break case with a 20% moisture content in the bicwdown.

The results of this analysis were a peak containment pressure of 47 psig and a peak temerature of 279'F. NNECO states that no AFWS flow was assumed in the criginal analysis based on operating procedures which require isolation cf the affected steam generater prior to manual AFWS initiation.

In NNECO's reanalysis, two containment pressure calculations were performed to envelope the effects of a single active failure upon the containment The most limiting single f ailure resulted from assuming pressure response.

f ailure of cne diesel generater with the: resultant loss cf one-half of the imortant ESF (one containment spray pump and two containment air recirculation fans). The AFWS pumo run-out flow used was 2050 gpm, based on a conservative backpressure of one atmosphere.

This reanalysis shows that the peak con-tainment pressure remains 47 psig if AFWS flow is delayed for three minutes.

It assumes the affected steam generator is not isolated resulting in a second increase of containment pressure up to almost 45 psig.

The staff concu s with the licensee's conclusion that the peak containment pressure will remain below the containnent design pressure after the MSLB accident with the addition of auxiliary feecwater at the ran-out flow rate three minutes af ter low steam generator level is reached.

Our review also included evaluation of the licensee's ability to determine and isolate the affected steam generator. NNECO states that the key parameter available to the operator following an MSLB would be low steam generator pressure in the affected steam generator. The MSLB analysis indicates automatic MSIV closure initiated at approximately three seconds after the break and a secondary side pressure of 500 psia (trip setpoint) in the affected steamThe generator versus approximately 695 psia in the intact steam generator.

mismatch becomes greater, approximately 98 psia in the affected steam generator versus 547 psia in the intact steam generator at 80 seconds after the break.

The plant operating procedures are written to enable a quick deter nination Once the determination of the steam line rupture and affected steam generator.

is completed approximately five seconds are required to manually isolate the affected steam generator stopping AFWS flow.

Based upon the above described control room indications, we find suffuient justification to assume the operator will be alerted to tne need to isolate the AFWS flow path to the affected steam generator before initiating AFWS flow manuall.y or within 10 minutes if automatic initiation is reflied upon.

2.3 Effects of Three Minute Delay of AFWS Flow on Other Transients and Accidents In addition to reviewing the effects of automatically initiating the AFWS in three minutes on the MSLB accident, we considered any adverse effects upon other transients and accidents. For example, assuming liquid discharge from a ruptured A

feedwater line, the reactor would lose one steam generator as a heat sink.

delay of AFWS injection could extend the heatup of the primary coolant system; however, the intact steam generator requires in excess of 10 minutes to boil dry and, therefore, provides an adequate heat sink for decay power removal.

Mf11 stone's Operating Procedures have historically required the initiation of AFWS as a manual action. Whenever credit for operator action was required, the i

analysis performed demonstrated the acceptability of the unit to withstand the j

postulated event being independent of operator action for a minimum of 10 I

minutes. We, therefore, conclude that automatic initiation of AFWS flow three I

minutes into the transient or accident (versus 10 minutes assuming operator l

action) is apprcpriate and would not result in consequences more limiting i

than previously analyzed.

2.4 AFWS Pumo Rated Capacity The Millstone-2 AFWS contains two electric motor-driven pumps, one supplied emergency power from each diesel generator, and one steam turbine-driven The pump with twice the rated capacity of either electric-driven pump.

l design objectives of the system, according to FSAR Section 10.4.5.3 is to provide feedwater for the removal of sensible and decay heat anti to cool the primary system to 300*F in case the main condenser and steam generator feed i

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pumps are inoperative due to loss of nonnal electric power sources or the The steam turbine-driven pump with a capacity of 600 gpm at main steam.

The electric-driven 2437 foot head was considered a " full-capacity" pump.

pumps with capacities of 300 gpm each at 2437 foot head were " half-capacity" rated.

20, 1980, NNECO provided a reanalysis to show that By application dated May the capacity of one electric motor-driven pump (300 gpm) was adequate to This was done meet the sensible and decay heat requirements to cool the RCS.

to justify automatically starting only the two electric motor-driven pumps.

The assumptions used were:

One of the electric motor-dr%2n pumps is inoperable; e

The turbine bypass to the condenser is unavailable; e

The AFWS pump starts 240 seconds af ter the stean generator low e

level is reached.

t This reanalysis shows the steam generator inventory decreases to a minimum of 9,400 lbs per generator, about 70 minutes from the reactor trip, before The steam generator inventory loss and recovery can be recovery starts.

The reanalysis improved by manually starting the steam turbine-driven pump.

further indicates the peak RCS pressure occurs about nine minutes into the event and that the PORVs (setpoint 2400 psi) will not be automatically lifted.

Our criteria has been "to automatically initiate AFWS flow". Following the criteria for other ESF systems, however, we agree with NNECO that installing a circuit to automatically initiate two independently powered electric motor-driven AFWS pumps, each rated at 100% capacity, is adequate to meet our short tern requirem NNECO finds some disadvantages to automatically initiating all three pumps.

First, they contend that normal unit startup and shutdcwn is routinely made with Starting all three pumps at one time auto-only one electric motor-driven pump.

Under these conditions of excess matica11y would provide 400f, of needed flow.AFWS flow, NNECO be l

within the limit specified in the TS.

Secondly, NNECO has noted cavitation of the Millstone t

in possible impeller damage.

25, 1980.

The final disadvantage was addressed in NNECO's letter of January They state that it was recognized that during a postulated steam generator tube rupture event, steam generator levels will drop to the l

j Al though This turbine exhausts directly to atnosphere.

l the Terry turbine.

use of the Terry turbine was previously a possibility during a steam generator l

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tube rupture event, the modification under consideration would result in an automatic start of the Terry turbine during this event. Therefore, NNECO perfonned a reanalysis of the radiological consequences of this event using the methodology previously documented, and assuming a release for thirty minutes from the Terry turbine exhaust. The results of this analysis are:

Whole Body Dose Thyroid Dose (Rem)

(Rem) e Results presented in Reference Cycle Analysis 0.006 0.1 e Additional Dose From Terry Turbine 0.048 0.003 e Total 0.054 0.1 Ve find that although the radiation exposure increase due to running the turbine-driven pump is not a significant increase from past analysis and is a small fraction of the dose guidelines of 10 CFR 100, added to the possi-bility of RCS overcooling and pump cavitation, there is a need }o carefully consider these factors in the design of the automatic start circuit for the turbine-driven pump.

We conclude that automatic initiation of the two electric motor-driven AFWS pumps meets our short term requirements. However, automatic initiation of the steam turbine driven AFWS pump is necessary to meet our October 22, 1979 requirement GL-3.

See Section 2.5.3 of this SE for an evaluation of GL-3.

2.5 NNECO Response to NRC Reconsnendations In response +w our letter of October 22, 1979, NNECO provided responses to our short-tem and long-tern recommendations by letters dated November 28, 1979 and January 17, March 10 and June 16, 1980. Our evaluation of these responses is as follows:

2.5.1 Short Term Recommendations Recorrnendation GS Emergency procedures for transferring to alternate sources of AM supply should be available to the plant operators. These procedures should include criteria to inform the operator when, and in what order, the transfer to alternate water sources should take place.

The following should be covered by the procedures:

The case in which the primary water supply is not initially avail-e able. The procedures for this case should include any operator actions required to protect the AFW system pumps against self-damage before water flow is initiated; and, The case in which the primary water supply is being depleted.

o The procedure for this case should provide for transfer to the siternate water sources prior to draining of the primary water supply.

Evalur'icn GS 4 - The licensee's response is acceptable in that applbble emergency procedures were revised to reflect the require-ment for transfer to alternate sources of AFW supply before January 1, 1980.

Recomendation GS The plant should be capable of providing the required AFW flow for at least two hours from one AFW pump train independent of any alternating current power source.

If manual AFW system initiation or flow control is required following a complete loss of alternating current power, emergency procedures should be established for manually initiating and controlling the system under these conditions.

Since the water for cooling of the lube oil for the turbine-driven pump bearings may be dependent on alternating current power, design or procedural changes shall be made to eliminate this dependency as soon as practicable. Until this is done, the emergency procedures should provide for an individual to be.

stationed at the *.urbine-driven pump in the event of the loss of all alternating current power to monitor pump bearing and/or lube oil temperatures. If necessary, this operator would operate the turbine-driven pump in an on-off mode until alternating current pcwer is restored.

Adequate lighting powered by direct current power sources and communications at local stations should also be provided if manual initiation and control of the AFW system is needed.

(See Recommendation Gl. 3 for the longer-tenn resolution of this concern.)

Evaluation GS-5.The licensee's response is acceptable. They have confirmed that emergency procedures were revised before January 1, 1980 to contain the information required to provide AFWS flow from one pump, independent of AC power source, for at least two hours.

Adequate portable lighting and communication are available for the prescribed manual actions.

Recommendation GS The licensee should confirm flow path avail-aDility of an AFW system flow train that has been out of service to perform periodic testing nr maintenance as follows:

L Procedures should be implemented to require an operator to e

determine that the AFW system valves are properly aligned and a second operator to independently verify that the valves are properly aligned.

e The licensee should propose Technical Specifications to assure that prior to plant startup following an extended cold shut-down, a flow test would be perfonned to verify the nonnal flow l

path from the crimary AFW system water source to the steam l

generators. The flow test should be conducted with AFW system valves in their nonnal alignment.

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. Evaluation GS The licensee's response is acceptable.

It was our posi:1on tnat procedures should be iglemented to require an operator to determine that the AFW system valves are properly aligned and a second operator to independently verify that the valves are properly ali gned. The licensee has agreed to perform this valve verification in accordance with our position. By letter dated March 31, 1980, the licensee proposed modifications to the TS to assure that prior to plant startup foll,owing an extended cold shutdown, a flow test would be performed to verify the normal flow path from the primary AFW system water source to the steam generators. The AFWS valves will be in their normal alignment during the flow test.

Recommendation GS The licensee should install a system to automatically initiate AFW system flow.

For the short-term, this system need not be safety-grade; however, it should meet the criteria listed below, which are similar to Item 2.1.7a of NUREG-0578. For the longer term, the automatic initiatiog signals and circuits should be upgraded to meet safety-grace reqdirements as indiested in Recomendation GS-1.

The design should provide for the automatic initiation of the e

auxiliary feedwater system flow.

The automatic initiation signals and circuits should be designed e

so that a single failure will not result in the loss of auxiliary feedwater system function.

Testability of the initiating signals and circuits should be a e

f eature of the design.

The initiating signals and circuits should be powered from the e

emergency buses.

Manual capability to initiate the auxiliary feedwater system e

l f rom the control room should be retained and should be igle-mented so that a single f ailure in the manual circuits will I

not result in the loss of system function.

The alternating current motor-driven pumps and valves in the e

auxiliary feedwater system should be included in the automatic actuation (sinultaneous and/or sequential) of the loads to the emergency buses.

The automatic initiation signals and circuits should,be designed e

so that their failure will not result in the loss of manual capability to initiate the AFW system from the control room.

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! The NNECO Evaluation GS The licensee's response is acceptable.

20, 1980 proposed to automatically start the two letter of May electrically driven AFW pumps and hold the turbine driven AFW pump as a manual backup. Their analysis shows that one electrically driven AFW pump (assuming failure of the second pump) will keep the steam generator level above the tubes sufficient to remove decay heat from the reactor coro. See Section 2.4 of this SE.

2.5.2 Additional Short-term Recommendations Recommendation 1 - The licensee should provide redundant level indications and low level alarms in the control room for the AFW system primary water supply to allow the operator to anticipate the need to make up water or transfer to an alternate water supply and prevent a low pump suction pressure condition from occurring.

The low level alarm setpoint should allow at least 20 minutes for operator action, assuming that the largest capacity AFWS pump is operating.

t Evaluation 1 - The licensee's response is acceptable as recently The installed level instrumentation in the condensate modified.

storage tank consists of low level alarm and a low-low level alarm in the control room. The licensee has confirmed that both the low and low-low level set points on the condensate storage tank are set to allow more than 20 minutes for operator action assuming that the largest capacity, AFWS pump is operating.

Recommendation 2 - The licensee should perform a 72-hour endurance test on all AFW system pumps, if such a test or continuous period Following the 72-of operation has not been accomplished to date.

hour pump run, the pumps should be shut down and cooled down and then restarted and run for one hour. Test acceptance criteria should include demonstrating that the pumos remain within design limits with respect to bearing / bearing oil temperatures and vibration and that pump room ambient conditions (temperature, humidity) do not exceed environmental qualification limits for safety-related equip-ment in the room.

Evaluation 2 - The staff has reduced the required performance test to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of operation instead of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The NNECO response is acceptable for the "B" motor driven and the steam turbine driven AFW Data provided in the NNECO letter of January 17, 1980 documents pumps.

reliable operation of these two AFWS pumps for considerable length The licensee should be required to conduct a 48-hour of time.

endurance test of the "A" motor driven AFW pump within 30 days from issuance of the SE.

Recommendation 3 - The licensee should implement the following requirements as specified by Item 2.1.7.b on page A-32 of NUREG-0578:

. Safety-grade indication of auxiliary feedwater flow to each steam e

generator should be provided in the control room.

The auxiliary feedwater flow instrument channels should be powered e

from tne emergency buses consistent with satisfying the emergency power diversity requirements for the auxiliary feedwater system set forth in Auxiliary Systems Branch Technical Position 10-1 of the Standard Review Plan, Section 10.4.9.

Evaluation 3 - The licensee's response is acceptable. NNECO has committed to provide safety grade AFWS flow indicator to each steam generator by July 1,1981, as required by NUREG 0737.

Recommendation 4 - Licensees with plants which require local manual realignment of valves to conduct periodic tests on one AFWS train, and there is only one remaining AFWS train available for operation, snould propose Technical Specifications to provide that a dedicated individual who is in communication with the control room be, stationed at the manual valves. Upon instruction from the control room, this operator would realign the valves in the AFWS train from the test mode to. its operational alignment.

Evaluation 4 - NNECO's response is acceptable. The licensee indicates tnat local manual valve realignment is not required for periodic testi.g.

2.5.3 Long-Term Recomendations Recomendation GL Licensees with plants having a nomal starting AFa5 snould install a system to automatically initiate the APdS flow.

This system and associated automatic initiation signals should be designed and installed to meet safety-grade requirements. Manual AFWS start and control capability should be retained with manual start serving as backup to automatic AFWS initiation.

Evaluation GL NNECO has installed the control grade circuitry required to automatically start the two electric motor-driven AFWS pumps. They have also committed to upgrade and replace components as necessary to meet safety-grade requirements. Our review of the safety-grade components will be completed and issued at a later date.

Recomendation GL At least one AFW system pump and its associated flow patn ano essential instrumentaion should automatically initiate AFWS flow and be capable of being independent of any alternating current power source for at least two hours. Conversion of. direct power to alternating current is acceptable.

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... Evaluation GL NNECO has stated that the upgrading of the turbine oriven AFW5, oy removing all AC dependence, will be completed by January 1,1982. We find this will improve the reliability of the turbine driven AFWS during a station blackout. However, during such a postulated event, the control room personnel could be under consider-able pressure to restore an electrical source thereby diverting their attention from the manual initiation of AFWS flow by starting the steam turbine-driven pump. Therefore, we continue to require ' automatic initiation of the steam turbine-driven AFWS pump in order to meet the criteria of specific Requirement GL-3 of our October 22, 1979 letter. This criteria does not precluds a design with special features to alleviate the disadvantages that NNECO identified in their letters of January 25 and May 20, 1980, as discussed in section 2.3 of this

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SE.

2.6 Technical Specifications Changes The proposed TS changes under review are from NNECO's applications dated December 27, 1976 and March 31, May 20 and August 29, 1980.

Some portions of the proposed TS changes or related TS pages should be modified to meet our requirements or for increased clarification.

Such modifications have been discussed with and agreed to by the NNECO staff.

Pages 3/4 3-13. 3/4 3-la, 3/4 3-15 and 3/4 3-17 Editoral change to correct numbers of action items.

l Pages 3/4 3-15, 3/4 3-20, 3/4 3-21 and 3/A 3/22 The automatic initiation of AFWS requirements should be added to the ESFAS Tables 3.3-3, 3.3-4 and 3.3-5.

Page 3/4 3-24 The surveillance requirements for automatic initiation of AFWS should be added to Table 4.3-2.

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... Pages 3/4 7 4 and 3/4 7-5 The surveillance requirements needed to prove the operability of the APWS should be expanded to include automatic initiation, flow path verification and valve alignment.

Page B 3/4 7-2 The AFWS pump basis should be changed to specify the new capacity rating of the electric or turbine-driven pumps.

3.0 Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this detemination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR $51.5(d)(4), that an environmental impact statement or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.

4.0 Safety Conclusion We have concluded, based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will De conducted in compliance with the Commission's reculations and the issuance of this amendment will not be inimical to'the common defense and security or to the health and safety of the public.

Date: January 14, 1981

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