ML19270H003

Forwards Proposed Tech Spec Changes Re Containment Purge Sys & Single Dropped Rod Protection.Safety Evaluations Encl

ML19270H003
Person / Time
Site:	Farley
Issue date:	06/20/1979
From:	Clayton F ALABAMA POWER CO.
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 7906220347
Download: ML19270H003 (18)

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tho toWan c4x!nc system June 20, 1979 Docket No. 50-348 Director cf Nuclear Reactor Regulation

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U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Mr. A. Schwencer Re: Changes to Operating License No. NPF-2 Technical Specificatioas

Dear Mr. Schwencer:

Alabama Power Company proposes the attached changes to Joseph M.

Farley Nuclear Plant Operating License No. NPF-2 Technical Specifications involving the following items:

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1. Technical Specifications 3.6.1.i, 4.6.1.7, B3/4.6.1.7, Table 3.3-5 and Table 3.6-1 concerning the Containment Purge System. The change adds the 18 inch mini-purge valves to the technical specification. The change also limits purging with the 48 inch purge valves to ninety (90) hours per year.

2. Technical Specifications B/2.2.1 and Table 2.2-1 con-cerning Single Dropped Rod Protection. To prevent possibly exceeding the DNB limit with the renctor in the automatic control mode, the negative flux rate trip will be decreased from 5 percent to 3 percent and the rate-lag time constant will be decreased from 2 seconds to 1 second. The positive flux rate trip rate-lag time constant will also be changed from .i seconds to 1 second.

Alabama Power Company has installed the containment mini-purge valves and plans to implement the Single Dropped Rod Protection at some future date. However, Alabama Power Company commits to operate with the reactor in the manual control mode until the change has been implemented. The analysis in the FSAR for the single rod drop event with the reactor in the manual control mode remains valid.

The Plant Operations Review Committee and the Nuclear Operations Review Board have reviewed the above proposed changes and have determined that the changes do not involve an unreviewed safety question as shown in the attached safety evaluation.

2348 056 79062201347.

n' Mr. A. Schwencer PAGE TWO June 20, 1979 These changes are deemed not to involve a significant hazard considera-tion, which is considered as a Class III change according to 10 CFR Part 170.22. A check for $8,000.00 is enclosed to cover the total amount of fees required.

In accordance with 10 CFR 50.30(c)(1)(i), three (3) signed originals and thirty-seven (37) additional copics of the proposed changes are enclosed.

If you have any questions, please advise.

Sincerely,

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'u A F. L. Clayton Jr.

FLCJr/ KAP:bhj SWORN TO Ah3 SUBSCRIBED BEFORE Enclosures METHIS)dfftDAY OF JUNE, 1979.

cc: Mr. R. A. Thomas

.. Mr. G. F. Trowbridge

1. < m '/ L& D NOTARY PUKLIC My Commission Expires: )'/ 5- k 2348 057

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ENCLOSURE SAFETY EVALUATION FOR TECHNICAL SPECIFICATION CHANGES ASSOCIATED WITH DROPPED ROD PROTECTION

Background:

Westinghouse recently notified Alabama Power Company that a review of safety analysis methodology for the single dropped rod indicated a potential for that event to lead to calculated DNB ratios lower than reported to the NRC for the Farley Nuclear Plant (FNP) class of plant. The single rod drop event is a DNB limited transient which is described in Section 15.2 of the FNP FSAR. The calculated consequences of the single dropped rod event are dependent upon whether the reactor is being operated in an automatic or manual control mode.

The impact of the inconsistency in the safety analysis methodology effects only the analysis of the single rod drop event with the reactor in the automatic control mode. The analysis in the FNP FSAR for the single rod drop event with the reactor in a manual control mode remains valid.

If a single rod drop event occurs when the reactor is in the automatic control mode, the reactor control system responds to both the reactor power drop (mis-match between turbine power and reactor power) and the decrease in the core average temperature and attempts to restore both quantitites to their original values. This restoration of reactor power by the reactor control system may result in some power overshoot depending upon the location of the excore power signal. In the case of the Farley Plant, the power signal is obtained from a single dedicated excore detector. Recent analyses by Westinghouse indicate -

that for a drcpped rod in the core quadrant adjacent to the dedicated excore detector, the power overshoot in the quadrant diagonally located from the dropped rcd is greater than the value calculated by the methods used in the FSAR. This could then 1ead to exceeding the DNB 1imit.

In the case of the FNP FSAR analysis for the single dropped rod, no credit is taken for the negative flux rate trip. Based on a recent Westinghouse analysis for the Farley Plant, it is proposed to change the high negative flux rate trip setpoints

- to assure that all dropped rod events result in a reactor trip. The negative flux rate trip would be decreased from 5 percent to 3 percent and the rate-lag time constant would be decreased from 2 seconds to 1 second. Since the electrical component generating this time constant also supplies the time constant for the high positive flux rate trip as shown in Figure 1, it is also proposed to change the positive flux rate trip rate-lag time constant from 2 seconds to 1 second.

References:

(1) Technical Specification Table 2.2-1 and Section B/2.2.1.

(2) FSAR Sections 7.2 and 15.2. 2348 058 Bases:

The negative flux rate trip setpoints should be changed as discussed above to provide additional assurance that a reactor trip will result as a consequence of

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a a',ngle dropped rod. This change is in a conservative direction from a safety standpoint and will assure that this transient will not result in a DNBR of less than 1.30. In addition, analyses show that the proposed changes will not result in spurious plant trips as a result of operational maneuvers and therefore should not affect plant reliability.

The change in the positive flux rate trip rate-lag time constant results in an increase in the positive rate required for reactor trip. The positive rate trip, however, provides a secondary means of protection for rapid power excursion transients (i.e., Rod Ejection Accident). Primary protection for these transients is provided by the neutron flux high/ low reactor trip. Thus, the accident analysis for these transients remains valid.

Conclusion:

The proposed changes to Technical Specification Table 2.2-1 and B/2.2.1 do not involve an unreviewed safety question as defined by 10CFR50.59.

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2348 059

FIGURE 1 LOGIC DIAGRAft FOR NEUTRON FLUX RATE TRIPS EICCEE 102 CitAM3ERS '

CONT AINING UPPER 1HD LOWER Parer ..

DETECTORS

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TEACTOR CORE MEASU'R ED WUtt.iAR POWER .

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FLUI R A T E FLUX ttSTAELE R a_iE E! STABLE

[ [ J CR CATE 4 _ . , _ _ j __ 4 10 PERMll C0VAan TRIP LOCtc ,

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CATE TO INITIATE i 2/4 TRi?

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TABLE 2.2-1 s REACTOR TRIP SYSTE!4 INSTRUf4ENTATION TRIP SETPOINTS A* ALLOWABLE VALUES Q FUNCTIONAL UNIT TRIP SETPOINT h 1. flanual Reactor Trip Not Applicable Not Applicable E

H 2. Power Range, Neutron Flux Low Setpoint - 1 25% of RATED Low Setpoint - 1 26% of RATED TilERf4AL POWER TilERf4AL POWER liigh Setpoint - 1 109% of RATED High Setpoint - 1 110% of RATED TilERMAL POWER TilERMAL POWER

3. Power Range, Neutron Flux, < 5% of RATED TilERMAL POWER with < 5.5% of RATED TilERMAL POWER liigh Positive Rate with a time constant > f seconds a 3time constant >fseconds 3

4. Power Range, Neutron Flux, < $% of RATED tiler L POWER with < f.5% of RATED TilERMAL POWER liigh Negative Rate a time constant _> seconds with a time constant _> / seconds f

5. Intermediate Range, Neutron 5 25% of RATED TilERMAL POWER 1 30% of RATED lilERf4AL POWER Flux 5 5

• 6. Scurce Range, Neutron Flux 1 10 counts per second 1 1.3 x 10 counts per second

7. Overtemperature AT See Note 1 See Note 3

8. Overpower AT See Note 2 See Note 4

9. Pressurizer Pressure--Low > 1865 psig > 1855 psig g _

u 10. Pressurizer Pressure--liigh 1 2385 psig i 2395 psig

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m 1 93% of instrument span

11. Pressurizer Water Level--liigh 1 92% of instrument span

-c3 O- > 90% of design flow > 89% of design flow

12. Loss of Flow _

per loop

• per loop *

• Design flow is 88,500 gpm per loop.

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2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Opera-tion with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

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Pov r Rance, Neutron Flux The Power Range, Neutron Flux channel high setpoint provides reactor core p otecticn against reactivity excursions which are too rapid to be protect.ed by temperature and pressure protective circuitry. The low set point frovides redundant protection in the power range for a power excursion beginning from icw power. The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of the four power range channels indicate a power level of above approximately 10 percent of RATED THERMAL POWER).

Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characte . tic of rod ejection events from any power level . Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.

2348 062 F

FARLEY - UNIT 1 B 2-3

O LIMITING SAFETY SYSTEM SETTINGS BASES

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Intemediate and Source Range, Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup. These trips provide redundant protection to the low setpoint trip of the Power Range, Neutron Flux channels +5 The Source Range Channels will initiate a reactor trip at about 10 counts per second unless manually blocked when P-6 becomes t ~ active. The Intermediate Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.

Overtemperature AT The Overtemperature AT trip provides core protection to prevent DN3 for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds), and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic com-pen'sation for piping delays from the core to the loop temperature detectors. With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1.

- If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

2348 063 FARLEY - UNIT 1 B 2-4

INSERT TO PAGE B 2-4 The Power Range Negative Rate trip provides protection to ensure that the minimum DNBR is maintained above 1.30 for rod drop accidents. At high power a single-or multiple rod drop accident could cause local flux peaking which when in conjunction with nuclear power being main-tained equivalent to turbine power by action of the automatic rod control system could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping

.the reactor for all single or multiple dropped rods.

2348 064

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ENCLOSURE 2 ,

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SAFETY EVALUATION FOR .

CHANGES TO TECHNICAL SPECIFICATIONS 3.6.1.7, 4.6.1.7, B3/4.5.1.7, TABLE 3.3-5 AND TABLE 3.6-1 Backcround:

The containment purge system for the Farley Nuclear Plant has the capability to purge the containment through 48 inch valves or 18 inch valves. The 18 inch valves were added to the purge system to be able to continuously purge the containment during plant operation. These valves should thus be added to the containment isolation valve list in the technical specifica -

tions. In addition, the response time requirements associated with mini-purge isolation should be added to the technical specifications. A technical specification for operation of the containment purge system utilizing the 48 inch ~ valves should be added to impose restrictions consistent with the plant safety analyses and NRC requirements.

References:

(1) NRC letter to Alabama Power Company dated November 28, 1978.

I (2) Alabama Power Company letter to NRC dated January 9,1979.

(3) rJAR Section 15.4.

(4) Alabama Power Company letter to NRC dated February 3, 1979.

. Bases:

The following provides the basestf or the technical specification changes by table and paragraph numbers.

Technical Specification Table 3.3-5 and Table 3.6-1:

. The containment is purged during normal plant operation by using the 18 inch mini-purge valves. These valves were designed in accordance with applicable codes and standards for containment isolation valves. This included designing for

- conditions (temperature, pressure, humidity, radiation) under which the valves must operate in the event of a LOCA. - .

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An analysis of the radiological consequences of a LOCA dur-ing operation of the mini-purge system was performed. This analysis resulted in incremental doses above that pre-sented in the FSAR LOCA analysis. These incremental doses, when added to the doses presented in FSAR Table 15.4-12, re-suited in the following doses.

2348 065

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-!. Enclosure 2 Page 2 Thyroid Dose, Rem

' 10CFR100 Table 15.4-12 Incremental Total Limits Site Boundary (2 hrs) 175 5.7 180.7 300.0 300.0 112.1 LPZ (0-30. days) 110 2.1 Whole Body, Rem 10CFR100 Table 15.4-12 Incremental Total Limits Site Boundary (2 hrs) 6.5 8.7 (10-3) 6.509 25.0

- LPZ (0-30 days) . 3.2 2.2 (10-3) 3.202 25.0 As seen from the above tabulation, the resultant doses are within 10CFP.100 limits and are therefore acceptable.

Additionally, an analysis of the reduction in the containment pressure resulting from the partial loss of containment at-mosphere during a LOCA was performed for ECCS backpressure de-termina tion. This analysis was based on the containment condi-tions defined in the limiting FAC amlysis case (DECLG break, CD = 0.4) obtained using the February,1978 Westinghouse Evalua-

,- tion fbdel. The impact on containment pressure resulting from this loss of containment atmosphere was less than 0.25 psi. The effect of a containment pressure reduction of this magnitude on the calculated peak clad temperature is expected to be minor (less than 200F). When added to the current LOCA peak clad temperature of 21580 F , the results of this evaluation indicate that the Farley

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Plant meets 10CFR50.46 limits (22000F) with the mini-purge system even if the containment is being purged at the time of a LOCA event.

In the event of a LOCA, these valves receive a signal to close as the result of containment high pressure. A 5-second closure time for these valves (including instrumentation delays) was assumed in the safety analysis. Therefore, these valves

. should be added to Table 3.6-1 with a 5 5.0 second closure time specified. In addition, a response time of s 5.0 seconds should be specified in Table 3.3-5 for containment mini-purge isolation.

I Technical Specification Paragraphs 3.6.1.7, 4. 6.1.7, and B3/4.6.1.7: '

The containment is purged during cold shutdown and refueling modes of plant operation by using the 48 inch main purge

_ valves. The 48 inch valves are also used for limited (not to exceed 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year) purging activities during plant operational modes. These valves were designed in accordance 2348 066

Enclosure 2 Page 3 .

with applicable codes and standards for containment isolati,on valves including conditions under which the valves must op-b'owever, an evaluation of the erate in the event of a LOCA. -.

- impact of purging during normal plant operation on ECCS per-forcance and the subsequent radiological consequences Therefore, Para- in the event of a LOCA has not been conducted.

graphs 3.6.1.7 and 4.6.1.7 should be added to incorporate the appropriate LCO and surveillance requirements for limited purg-The correspo ing utilizing the 48 inch main purge valves.

bases should be added in Paragraph B3/4.6.1.7.

I Conclusio n: '

The proposed changes to Technical Specification Table 3.3-5 ' ara 3.6-1 and unreviewed

,' Paragraphs 3.6.1.7, 4.u.1.7, and B3/4.6.1.7 do not involve safety question as defined by 10CFR50.59, 2348 067 O

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TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATitG SIGNAL A"O FUNCTION RESPONSE TIME IN SECONDS

1.  ::arnal

a. Safety Injection (ECCS) flot Applicable feedwater Isolation Not Applicable Reactor Trip (SI) I;ot Applicable Containment Isolation-Phase "A" Not Applicable Containment Vent and Purge Isolation Not Applicable Auxiliary Feedwater Ptaps Not Applicable Essential Service Water System Not Applicable Containment Air Recirculation Fan Hot Applicable .

Containment Spray Not Applicable b.

Ce7tainmant Isolation-Phase "B" Not Applicable t

' Containment Vent and Purge Isolation Not applicable Containment Isola tion-Phase "A" Not Applicable

_ c.

Containment Vent and Purge Isolation I;ot Applicable

d. Steam Line. Isolation Not Applicable

2. Containment Pressure-Hich '

a. Safety Injection (ECCS) 5,27.0*

b. Reactor Trip (from SI) 1 2.0

c. Feedwater Isolation 1 32.0 y .

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d. Containment Isolation-Phase "A" < 17.0 /27.0""

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&n Not Applicable

c. Containment Vent and 3 Purge Isolation Auxiliary Feedwater Pumps flot Applicable BR f.

Essential Service Water System 1 77.0 /87 0""

/ g.

h. Containment air Cooler Fan $_27.4

1. Contar'n ment hk'-/Bqc Infatah 4 .S*.o .

2348 068 ehan e # '

. FARLEY - UNIT 1 3/4 3-27

TABLE _3.6-1 (Cont'd)  :

Isolation

Phase A Isolation Function Timei~sec)

E (Cont'd) Train ,

<10 g

B CCW from exc letdown RCOT HXS <10

, 32. CCW-HV-3067 Accumulators fill line isolation B 10 e . 33. CVC-HV-8860 Accumulator tanks sampic isolation valve A <10 5 34. SS-HV-3766 Accumulator tanks sample isolation valve

-4 35. SS-HV-3334 B <10 A RCOT vent line isolation valve <10

36. LWP-HV-7126 8 RCOT vent line isolation valve .:10

37. LWP-HV-7150 Containment sump recirculation valve A -10

38. LWP-HV-3330 Demineralizer water to reactor HD storage '

39. CTS-HV-3659 B Containment purgegdsolation valve g g a u 9.. _5 B 5

40. CBV-HV-3196 Containment purge Supply 4v alve ,isolet%n 8

41. CBV-HV-3197

• INSERT 1 ...................

B. Phase B Isolation

<l5 R ^ A CCW to RCP coolers <l5

1. CCW-MOV-3052

2. CCW-MOV-3046 B CCW from RCP oil coolers <l5

? A CCW from RCP oil coolers <10 5' 3. CCW-MOV-3182 B CCW from RCP THRM BARR <10

4. CCW-HV-3184 CCW f rom RCP THRM BARR A <10

5. CCW-HV-3045 Containment instrument air supply valve A .

6. IA-HV-3611

'C. Safety Injection Signal

, <10 A Charging pumps to regenerative HX <10

1. CVC-MOV-8107 Charging pumps to regenerative HX

2. CVC-MOVa8108 8 <ls B SW to RCP motor air coolers <l5-

- 3. SW-MOV-3135 SW from RCP motor air coolers A <l5 . ,

j 4. .SW-MOV-3131 SW from RCP motor air coolers B

rs; 5. SW-MOV-3134

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I?lSERT 1 HV-2866 A Containment Mini-Purge Supply Isolation Valve 15

42. Containment Mini-Purge Supply Isolation Valve 15 HV-2866 B 43.

A Containment Mini-Purge Exhaust Isolation Valve 15

44. HV-2867 Containment Mini-Purge Exhaust Isolation Valve 15

45. HV-2867 B U

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CONTAIRIENT SYSTDIS .

CONTAIRIENT VENTILATION SYSTDi ,

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LU!ITING CONDITION FOR OPERATION 3 . 6 .1. 7 The 48 inch contain=ent purge supply and exhaust isolation valves (CBV-lW-3198A, 31933, 3196, 319 7) shall be:

(a) closed or,

- (b) open for a time period not to exceed 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year.

APPLICABILITY: MODES 1, 2, 3 and 4 ACTION:

With one containment purge supply and/or one exhaust isolation valve open, except as specified by 3.6.1.7 b r.aintain at least one isolation valve closed in each affected penetration and either:

(1) Close the open valve (s) within one hour, or (2) Isolate each affected penetration within one hour by use of at least one deactivated auto-natic valve secured in the isolated position, or (3) Be in at least IIOT STED3Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD S11UTDOWN within the follow-ing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIRDIENTS 4.6. l. 7 At least once per 31 days:

(a) Determine that the containment purge supply and exh2ust isolation valves are closed, or p) Verify that the valve (s) is open as permitted in 3.6.1.7 b.

2348 071 9

D FARLEY-UNIT 1 3/4 6-10A e

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C0tiTAll!MEllT SYSTEMS BASES The maximum peak pressure e>.pected to be obtained from a LOCA event is 45 psig. The limit of 3 psig for initial positive containment pressure will limit the total pressure to 48 psig which is less than -

design pressure and is consistent with the accident analyses.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that the overall containment average air temperature does not exceed the initial' temperature condition assumed in the accident analysis for a LOCA. ,

3/4.6.1.6 C0fiTAIt? MENT STRUCTURAL If1TEGRITY This limitation ensures that the structural integrity of the contain-

' ment will be maintained comparable to the original design standards for --

the life of the facility. Structural integrity is required to ensure (

that the containment will withstand the maximum pressure of 48 psig in"' ~

the event of a LOCA. The measurement of containment tendon lift off force, the visual and physical examination of tendons, anchorages and liner, and the Type A leakage. test are sufficient to demonstrate this ,

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canabilit 3/4.6.2 DEPRESSURIZATI0tl At:D C00LIfiG SYSTEMS 3/4.6.2.1 CONTAltiMEt!T SPRAY SYSTEf1 The OPERABILITY of the containment spray system ensures that contain-ment depressurization and cooling capability will be available it, the event of a LOCA. The pressure reduction and resultant lcwer containment leakage rate are consistent with the assumptions used in the accident analyses.

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- 23'48 072 f% .

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FARLEY-VillT 1 8 3/4 6-2 .

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IllSERT 2 -

3/4.6.1.7 C0t!TAltNEllT vet 1TILATIO?l SYSTEM The 48" containment purge supply and exhaust isolation valves have been demonstrated capable of closing during.a LOCA. However, an evaluation of the impact of purging during plant operation on ECCS performance and the subsequent radiological consequences in the event of a LOCA has not been conducted. Therefore, these valves are not permitted to be open during plant operation for more than 90 .

hours per year. .

4 2348 073 O

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