ML19347C848
| ML19347C848 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 02/20/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19347C847 | List: |
| References | |
| NUDOCS 8103100086 | |
| Download: ML19347C848 (9) | |
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O ENCLOSURE FORT CALHOUN SER ON AUTOMATIC INITIATION OF AUXILIARY FEEDWATER A '.
Introduction On October 30, 1979, a letter was sent by Mr. Harold Denton, NRC,to all operating nuclear power plants. The subject of this letter was entitled,
" Discussion of Lessons Learned' Short Term Requirements." Enclosure 1, in this referenced document, describes NRC's requirements for upgrading the reliability of the auxiliary feedwater system (AFWS).
Included among these requirenents are the following:
(a) The auxiliary feedwater system shall have automatic / manual initiation capabilities, and; (b) The failure of either circuitry (automatic or manual controls) will not affect the reliability and operability of the other circuitry.
Also, in a letter dated December 21, 1979, the NRC staff stated its concern regarding the applicability of current analyses for main steam line breaks
.(MSLB) under conditions of automatic initiation of auxiliary feedwater, and the staff requested that the licensee, Omaha Public Power District (OPFD),
address this concern.
IE Bulletin 80-04 dated February 8,1980, provided additional guidance to the licensee regarding reanalysis of the MSLB.
In response to the above letter and IE Bulletins,' 0 PPD subnitted Reference 1 which describes plant modifications and analyses to support compliance with the staff's requirement to automate the AFWS. Historically, Fort Calhoun has relied upon operator action to initiate the AFWS. The NRC's Lessons
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r 2-Learned Task Force (NUREG-0578) detemined that plant safety can be signifi-cantly improved by insuring the automatic initiation of AFWS, thus pro-viding the cperator additional time to assess the status of the plant under abnomal conditions.
During our evaluation of Fort Calhoun's =cdified auxiliary feedwater systen, the following major concerns were addressed:
(a) The effect of a three minute delay in initiation of the AFWS during MSLB; (b) The licensee's consideration of single failure in the MSLB analysis; i
(c) The implications of loss-of-offsite power (LOOP); and, (d) The implications-of the three minute delay in AFWS initiation on other transients and accidents.
Each of the above items is addressed below.
B.
MSLB Accident - Return to Power OPPD's analysis of the effects of return to power following a MSLB accident
. is presented in Reference 1.
The starting conservative asstt ptions, according to OPPD, for this analysis are:
Only a three-minute delay in delivery of auxiliary feedwater flow to the steam generator was assumed, rather than a more realistic longer time delay accounting for tne celay in AFWS signal initiation and the transit time of the feedwater flow to the steam generator,
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.. Credit :7 not taken for complete isolation of the main feedwater system, thereby resulting in a continuous flow of 5 percent of full flow of main feedwater to the affected steam generator, A conservative high auxiliary feedwater flow was assumd to be fed entirely to the damaged steam generator, Failure of one HPSI pump, Failure of one LPSI pump, The highest worth CEA is assumed to stick in the fully withdrawn position
- and, The end of life moderator temperature and Doppler (fuel temperature) coefficient values were used since these values result in the greatest positive reactivity change during cooldown.
The analysis assts:ed that the event is initiated by a circumferential rupture of a 34 inch main steam line at the steam generator nczzle. GPPD states that this break is limiting since i,t results in the greatest rate of temperature reduction in the reactor core region. Reactor trin and safety injection follow the pipe rupture. This reanalysis reported in Reference 1 uses the same assu=Dtions and methods as previously used in the FSAR and subseouent licencing submittals except that it siculates au+w-matic initiation of auxiliary feedwater flow in three minutes from initiation of the event.
The rationale for delaying the initiation of AFWS originates from the cositive reactivity feedback which accompanies a postulated MSLB. During a postulated
. double-ended guillotine break of this steam lin'e, the broken steam generator behaves as an enhanced heat sink, resulting in rapid cooldown of the primary system. This rapid cooldown has a noticeable impact on the moderator reactivity feedback, which results in a net positive reactivity inserthn.
A conservative assumption is made that the limiting control element assembly (CEA) is stuck in its fully withdrawn position.
Based on s..e licensee's generic analyses, the reactivity feedback was most limiting for a main steam line break during full power operation. For this condition, the calculations predict that there would be a 10.8% return to power due to the coolino effect of the auxiliary feedwater, however, this is less than the 12% return to power predicted prior to auxiliary feedwater injec-tion. The net energy removed from the primary system was conservatively assumed to be the product of the total steam generator mass (M TOT) times the latent heat of evaporation (hfg). Should liquid entrainment exit the break, then the energy removed from the primary system will be less severe.
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For a postulated guillotine break in a steam line, the time required to l
deplete the broken steam generator secondary inventory is approximately 70 seconds (for the full power condition). When the auxiliary feedwater is injected into the steam generator, the magnitude of the primary side is increased). This results in enhancing the l
.cooldown is increased (MTOT The mechanism primary side coolinj and in an increased reactivity feedback.
available for reversing the reactivity insertion is the initiation of the ECCS, l
f which injects boren into the system.
The licensee's assessment of the effects of automatic initiation of AFWS l
during' a postulated MSLB concluded that a three-minute ' delay in the initiation of AFWS will ensure that the previous analyses for MSLB submitted in 'the FSAR and subsequent reloads are conservatively bounding. The purpose
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, of the three-minute delay is to provide time for-the ECCS injected borated water to lessen the magnitude of the moderator reactivity feedback attributed to the AFWS inventory.
The licensee's analytical metnod for analyzing steam line breaks is presently under staff review. The review at this time indicates reasonable assurance that the conclusions based on the submitted analyses will not be appreciably altered by the completion of the analytical methods myiew.
The staff finds the return to 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, therefom, acceptable.
C.
MSLB Accident - Single Failure Considerations OPPD states that single failures concurrent with the MSLB, other than those listed in the assunptions, were not considered since any other postu'ated failures have not been part of the design basis as described in the FSAR.
While not directly relevant to staff approval of the automatic actuation of AFW, the licensee's vulnerability to sing 12 failures has been examined because new licensing analyses were submhted. Our conclusion is that although the licensee has not documented a complete evaluation of potential single failures, sufficient conservatism exists in the analyses for Fort Calhoun.
In particular, the licensee has included in the analysis assumptions the failure of the safet" grade MFW isolation valves as well as one HPSI pump. While the licensee has not addressed the failure in the open position of relief or steam dung valves loct.ted on the intact steam generator, generic analyses of MSLB for similar PWRs have indicated that the worst single failure is
. the loss of a HPSI pump as was assumed in the licensee's presert analysis.
Lastly, the SEP review of tne Palisades plant is to address single failures for the MSLB in greater detail. The review at this time indicates reasonable assurance that the licensee has adequately accounted for single failure. We will factor in the SEP results for Palisades at the conclusion of the overall program.
D.
MSLB Accident - Effects of loss of Offsite Power The licensee has not submitted calculational results concerning the ccnsequences of losing offsite power during a postulated steam line break.
The primary consequences resulting from loss-of-offsite power (LOOP) are a delay of emergency core cooling systems (ECCS) injection and tripping of the reactor coolant pumps. During LOOP, ECCS injection is delayed approximiltely 25 seconds as the emergency diesel generators restore power to the ECCS pumps. LOOP also results in coastdown of the reactor coolant pumps.
Continued operation of the reactor coolant pumps would have two effects on an SLB transienti Running the reactor coolant pumps (RCDs) results in a greater degree of overcooling as the hot primary fluid is forced through the steam generators, and The reactor coolant pumps act as a driving head, forcing the ECCS injected borated water into the core.
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. Thus, losing offsite power affects the degree of system cooling and the time at which the ECCS-injected boron enters the reactor core. Overcooling and borated water injection are conpeting effects in which the fomer increases reactivity and the latter reduces reactivity. While the licensee has not addressed the effect of LOOP on MSLB, in reviewing past analyses of MSLB for other plants similar to Fort Calhoun, we have determined that LOOP results in the injected boron tdf6g dominant over the RCS cooldown.
Thus, we find that'the reactivity effects of a MSLB accident would be reduced in severity with a concurrent loss of offsite power.
E.
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 feedwater line, the reactor would lose one steam generator as a heat sink. A delay of AfWS injection could extend the heatup of the primary coolant system; however, the intact steam generator requires in excess of ten. minutes to boil dry and, therefore, provides an adequate l
heat sink for decay heat removal.
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Fort Calhoun's Cperating Procedures have historically required the initiation of AFWS as a manual action. Whenever c. edit for operator action was required, l
the analysis perfomed demonstrated the acceptability of the unit to withnand the postulated event being independent of operator action for a minimum of ten minutes. We, therefore, conclude that automatic initiation of AFWS i
flow three minutes into the transient or accident (versus ti.n minutes assuming I
operator action) is appropriate and would not result in consequences more limiting than previously analyzed.
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.g.
F.
Conclusions he have reviewed the licensee's submittals addressing automatic initiation of the AF4S and conclude that the proposed method for auto atic initiation and supporting analyses are acceptable. The staff plans to per'orm independent audit calculations to provide further confirmation of our conclusions. We will request that the licensee provide appropriate input data for the audit calculations, by the identification of the specific data needed, in the near future.
G.
_Re fe ences 1.
Letter from W. C. Jones, OPPD, to NRC, Main Steam Line Break and Autocatic Initiation of AFWS, January 10, 1983.
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