ML19259B639
| ML19259B639 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 02/08/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19259B638 | List: |
| References | |
| NUDOCS 7903130486 | |
| Download: ML19259B639 (12) | |
Text
[
UNITED STATES yy J.
NUCLEAR REGULATORY COMMISSION
$ $1j[ )
WASHINGTCN, D. C. 20555
%gj,j s...e SAFETY EVALUATION BY TH" 0FFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 48 TO FACILITY LICENSE NO. " R-33 TENN_SSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT NO.1 DOCKET NO. 50-259 1.0 Introduction By letter dated September 8,1978, (TVA BFNP TS115), as supplemented by letters dated October 5,1978, November 30, 1978, December 5,1978, December 14, 1978, January 8,1979 and January 9,1979, the Tennessee Valley Authority (the licensee or TVA) requested changes to the Technical Specifications (Appendix A) appended to Facility Operating License No. DPR-33 for the Browns Ferry Nuclear Plant, Unit No.1.
In support of this reload application for Browns Ferry Unit No.1 (BF-1), the licensee submitted a reload licensing document prepared by the General Electric Conpany (GE), a supplemental reload licensing document also prepared by GE and proposed changes to the Technical Specifications.
On January 17, 1979, we issued Amendment No. 47 to Facility License No. DPR-33 in response to the above submittals. The amendment authorized TVA to startup and operate BF-1 in the third fuel cycle.
In the staff's safety evaluation accompanying this amendment, we evaluated the items requiring attention during reload reviews, including nuclear and mechanical design of the fuel, thermal-hydraulic, transient and accident analyses, startup testing programs, etc.; all staff concerns were satsifactorily resolved with the licensee. As discussed, below, the staff had' reservations about the testability of the end-of-cycle' recirculation pump trip (EOC RPT) feature which the licensee installed in BF-1 during the refueling outage.
(The E0C RPT is different from and should not be confused with the recirculation pump trip feature which has been installed in several boiling water reactors to mitigate the consequences of anticipated transients without scram; the latter is commonly refered to as ATWS RPT). The staff's reservation aoout the testability of the EOC RPT was resolved by not including credit for the 7903130486 E0C RPT system in. the operating limit minimum critical power ratios (OLMCPR) incorporated into the Technical Specifications with Amendment No. 47.
(See section 3.2.2 of the accompanying safety evaluation).
On January 17, 1979, a meeting was held with representatives of TVA and the General Electric Company (G.E.) to discuss the electrical design and testability of the EOC RPT. The information presented at this meeting resolved the staff's reservations on preoperational, startup and periodic testing of the EOC RPT system.
23,1979. (l)information was This documented in TVA's letter of January Accordingly, this amendment revises the. OLMCFRs to provide credit for the E0C RPT as justified by the analyses submitted by the letters referenced above.
2.0 Discussion Section 14.5 of the Browns Ferry Nuclear Plant Final Safety Analysis Report (BFNP FSAR) discusses the analyses of abnormal operational transients. The events that could result in significant nuclear system pressure increases are those that result in a sudden reduction of steam flow while the reactor is operating at power.
These possible events are:
(1) generator trip, (2) loss of condenser vacuum, (3) turbine trip, (4) turbine bypass valve malfunction, (5) closure of main steam isolation valve and (6) pressure regulator malfunction.
During the refueling outage (tbvember 26, 1978 to January 17, 1979),
TVA installed an end-of-cycle RPT system (hereafter referred to as RPT systert) in BF-1.
This system provides automatic trip of both recircula-tion pumps after turbine trip or generator load rejection if reactor power is above approximately 30 percent of rated full load. The purpose of this trip is to reduce the peak reactor pressure and peak heat flux resulting from trar.sients in which it is postulated that there is a coincident failure of the turbine bypass system. The recirculation pump trip signal results from either turbine control valve fast closure or turbine stop valve closure. Reactor scram is also initiated by.these signals.
Since the recirculation pump trip involves opening of circuit breakers between the motor-generator set and the pumps, the flow coastdown is more rapid than that resulting from loss of power to the motor-generator sets. The very rapid reduction in core flow following a recirculation pump trip early in these transients reduces the severity of the events because the immediate resultant increase in core voids provides negative reactivity which supplements tne negative reactivity from control rod scram.
. 3.0 Evaluation 3.1 Operatina limit Minimum Critical Power Ratio (0LMCPR)
Various transient events will reduce the MCPR from its normal operating value. To assure that the fuel cladding integrity safety limit MCPR will not be violated during any abnormal operational transient, the most limiting transients have been reanalyzed by the licensee to cetermine which event results in the largest reduction in critical power ratio.
Each of the events has been conservatively analyzed for each of the several fuel types (i.e., 7x7, 8x8, 8x8R) and for the full range of exposure through the cycle.
The methods used for these calculations, including cycle-independed initial conditions and transient input parameters are described
,1 Reference 2.
Our acceptance of the values used and related -
sient analysis methods appear in Reference 3.
Supplemental cycic Jancent initial conditions and transient input parameters used in tr.. analysis appear in the table in Section 6 and 7 of Reference 4.
Our evaluation of the methods used to oevelop these supplementary transient input values have already been addressed and appear in Reference 3.
The overall transient methodology, including cycle-indepenoent transient analysis inputs, provides an adequately conservative basis for the determination of transient ACPRs. The transient events analyzed were load rejection without bypass, turbine trip without bypass, feedwater controller failure, loss of 100*F feedwater heating and control rod withdrawal error.
In the analysis of these reactor transients, the licensee has proposed to take credit for an EOC recirculation pump trip (RPT).
This reduces the transient t.CPR during reactor core pressurization events (e.g., load rejectior., turbine trip) by tripping breakers in the electrical line between the motor-generator sets and the-recirculation pumps on closure of turbine stop or control valves.
The prompt RPT immediately reduces core flow and thereby increases core voids.. The innediate voiding provides negative. reactivity which supplements scram reactivity.
In this manner, the RPT re-duces the thermal power spiking during the pressurization events.
This RPT feature is a margin improvement option which was not generically approved in our evaluation of the reference reload topical.(3)
, The KPR credit for the prompt RPT was calculated with the REDY code. The REDY code employs a two node steamline thermal hydraulic model and a point kinetics neutronics model. Several pressurization experiments at Peach Bottom Unit 2 (Reference 5) were designed to check the validity of these REDY models.
The experimental results showed that trie REDY steamline model did not accurately predict pressurization rate which is the mechani::m reducing the CPR.
Also, the REDY point kinetics model could not simulate the axial reactivity variation in the core. GE imnadiately provided calculational comparisons of REDY and test results, and attempted to demonstrate that although REDY did not accurately model some transient effects, it did provide a conservative basis for current licensing calculations. -
We agreed with GE's general conclusion that REDY provides a conserva-tive calculation for the current licensing basis transients on oper-ating reactors. However, we also recognized that REDY's inability to accurately predict pressurization rate and axial reactivity re-sponse limits simulation of effects of RPT. The Peach Bottom tests demons a d 'he existence of a pressure wave phenomenon in the steam lines.
In addition, it was noted that the power rise asso-ciated with pressurization was significantly greater in the upper portion of the core than in the lower portion.
Quantitative comparison of the tests with REDY calculations indi-cated that the REDY model underpredicted the presrurization rate
.but overpredicted the core's response to pressurization effects.
Thus, there are two discrepancies between REDY simulated effects and real transient's effects. One is non-conservative and the other is conservative.
It is impossible to state from these com-parisons which effect would predominate for a given transient.
After the analysis of the tests, comparisons were made between REDY simulations and simulations using detailed steamline modeling j
and a time-varying axial power cistribution.(8) These comoari-i sons, altnough rather limited, suggest a trend in which REDY-based calculations conservatively predicted EPR for more severe tran-sients but underpredict aCPil (for a given set of input parameters) for less severe transients.(8) These calculations also showed that the aCPR benefits for the RPT feature may be overpredicted by REDY as compared to the cetailed steamline and core modeling predic-tions.
. In the face of this information, we decided to take no action for three reasons:
(1) operating limit MCMs are always based upon the most severe transient for each fuel type, (2) these limiting transients were sufficiently severe to te in the range where REnY-based calculations are conservatis., and (3) GE was developing a more sophisticated transient simulator to accurately predict the questioned phenomena.
However, with the addition of che RPT feature, the limiting pressure and power increase transier.c analyses generally predict a aCPR in the range where REDY is less conservative.
We find that full credit for the RPT effect cannot be justified solely on a REDY analysis.
Two alternatives suggest themselves as means of resolution. The first alternative is to provide additional justification for the proposed specification. The GE ODYN code has more nodes to model steamline dynamics than REDY and also has a one-dimensional axial core neutronics model. ODYN's develop.. ant has been based on first principles and verified by the Peach Bottom tests.
ODYN is currently under a staff review that is to be complete within the next few months. ODYN will be used as the calculational model for pressurization events when it is approved.
Prior to its approval, we find tnat ODYN could be used to simulate the RPT effects and, thereby, provide assurance of the ACPR benefit.
During this time, we will accept the greater aCPR of tne ODYN and REDY calculations.
Once ODYN receives generic approva~., we will accept the ODYN calculated ACPR regardless.
We are not reouiring an ODYN calculation, however. We have made it clear that we will evaluate any other justification which the licensee submits and all applicable calculations and data which become avail-able to us' through other channels.
Another alternative to an adequate aCPR for RPT effects is to con-servatively bound the REDY calculation.
In a previous reload safety evaluation for Browns Ferry Unit 3,( 9.) we found that an i.. crease of 0.05 in ACPR for the limiting pressure and power increase tran-sients or an increase of.07 in aCPR for the feedwater controller 1
f ailure transient (whichever is more limiting) will bound the poten-t tial non-conservatism in the analyses.
In the subsequent reload amendment,(10) operating limit MCPRs for only the initial 2000 mwd /t of Cycle 2 were provided, with the statement that MCPRs to ena of cycle will be determined by reanalysis. This position was reached from the licensee's insistence to not implement the conservatively adjusted staff required MCPR and the licensee's preference to rely on a timely GE core specific ODYN reanalysis with staff review.
Since this amendment was issued, we have had several telephone con-versations and me'etings with TVA. From these discussions, we have concluded that the increase in ACPR may De somewhat reduced. This conclusion is based on the fact "that the previously established
'ncrease was in part established from ODYN comparisons to REDY
'or measured scram times. These calculations would emphasize the
- t. ore axial modeling cifferences between the codes and, thus, enhance the differences by virtue of the phenomenon modeling characteristics.
With this, the ODYN-REDY comparison for the measured scram time can be somewhat de-emphasized.and the 4CPR differences between ODYN and REDY for the RPT can be considered to provide the primary indication of the effe'ct of the "more sophisticated" Toeling. The only avail-able comparison of ODYN and REDY for the RPT shows a oCPR difference of about 0.02.
This calculation is for a specific BWR which is different in plant size and core loading than the Browns Ferry Units.
On these bases, we and the licensee have agreed tnat a conservative bound to the REDY calculation with RPT would be assured with a 0.03 ACPR increase for rapid pressurization transients.
Based on our composite review of the licensee's submittals anc D.03 MCPR increase, the most limiting abnormal operational transient for all fuel types and exposure intervals except for the 7x7 fuel from BOC to EOC-2000 MWa/t is the load rejection without bypass. For the 7x7 fuel from BOC to E0C-2000 mwd /t, the limitin transient is the loss of 100*F feedwF.er heating.
The overating limit MCPRs which the licensee has JroposedOI) and which are acceptable to the staff are as follows:
OPERATING LIMIT MCPR Fuel Type E0C-2000 mwd /t EOC 7x7 '
1.20 1.25 8x6 1.24 1.30 8x8R 1.24 1.30 Thus, when the reactor ~ s operated in accordance with the above operating limit MCPRs the 1.07 SLMCPR will not be violated in the event of the most severe abnormal operational transient. This is acceptable to the staff.
3.2 Failure of Trio Inputs from Turbine Buildine to Reactor Protection System During our review.of the reactor protection system, we noted that the trip inputs for the recirculaticn pump trip and reactor scram following load rejection or turbine trip originate in ^.he turbine buildi ng. The turbine building, as is the case of mot,t boiling water reactor plants, is not seismically qualified; hence, its integrity and functions cannot be assured in the event of an ea rthquake.
For these reasons, the licensee was requested to analyze the conse-quences of a safe shutdown earthquake concurrent with the limiting transient event without taking credit for reactor scram or recircu-lation pump trip from the turbine building inputs. The licensee has-referenced generically applicable analyses.
We agree with the li-censee that this analysis is applipable and on the basis of previous staff findings on this analysis,(
) we find the results acceptable.
3.3 Electrical Centrol and Instrumentation Ascects of EOC RPT The design philosophy fer the EOC RPT system is described in GE report NEDO-24119, " Bas ; for Installation of Recircyla} ion Pump Trip System", Browns Ferry Nuclear Plant, April 1978tl3i. This report provides the safety evaluation of the instrumentation and control aspects of the proposed modification. The design of the RPT system was evaluated against the criteria of IEEE Standard 279.
The RPT feature serves as an essential safety supplement to the scram system and, as such, is required to comply with IEEE Standard 279.
Basically, the RPT feature consists of hydraulic pressure switches (sensors to detect the fast closure of the turbine control valves),
position switches (sensors to detect closure of the turbine stop valves), relays, logic, and fast-acting circuit breakers (actuation devices).
In order to satisfy the single failure criterion, the RPT logic feature consists of two almost-identical systems in a one-out-of-two configuration such that either is capable of operating independent circui' breakers in the supply circuit of each recirculation pump motor.
. The operation of any RPT sensor (pressure switch or position switch from any of the four turbine control valves or any of the four turbine stop val is) causes an electromagnetic relay to de-energize.
The relay contacts are combined with contacts from an Operating Bypass and contacts from a manual bypass switch to provide power to the breaker trip coils. The turbine valve sensors and turbine pressure sensors for the RPT feature are the same ones used for the scram system.
The Operating Bypass disables the RPT system when the turbine first-stage pressure is below about 30%, as is done for the turbine inputs to the scram system.
The manual bypass switch ("out-of-service") allows' each RPT system to be disabled for maintenance purposes.
A fast closure sensor from each of two turbine control valves provides input to one RPT system; sensors from the other two turbine control valves provide inputs to the second RPT system. Similarly, a position switch for each of two turbine stop valves provides input to one RPT system; sensors from the other two st;p valves provide imputs to the other RPT system.
For each RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of control valves and a 2-out-of-2 logic for the stop valves. The operation of either logic will actuate the RPT feature.
Information initially provided by GE and TVA indicated that the RPT feature should function when scram occurs for either of the two turbine transients.
However, the criterion that was finally established by GE and TVA requires the RPT to function only if all four turbine stop valves or all four control valves close. That is, if only three of the stop valve
_ilure within the RPT will prevent the RPT from providing system-level protective action in accordance with the final criterion.
TVA has stated that the RPT equipment will be appropriately qualifled as Class TE. We note that inputs to the RPT originate in the turbine building which is not a seismic category 1 structure.
The input equipment is adequately qualified for the anticipated occurrences of the turbine. As discussed in Section 3.2 above, we have evaluated this aspect of the design with respect to the potential consequences of a safe shutdown earthquake and conclude that there will be no undue risk to %e public health and safety from this postulated event.
9 The two systems of the RPT feature will be physically and electrically independent.
There is one interconnection between the RPT and a non-safety system. When the RPT is tripped, cuxiliary relay contacts feed the control circuits of the motor-generator sets to deenergize them.
This interlock is adequately isolated such that no credible failure can prevent proper action of the RPT.
Althcugh the purpose of the FST is to =iti ate a cere-wide pressuricatien 6
transient, the desired thermal margin advantage can be realiced only i' the i-itiating events are sensed en an anticipatery basis, rather than menitoring pressure directly.
The use of pressure switches te sense the 1 css of hydraulic cent ol fluid pressure to' each control valve is adecpate v.' 'k..=.=.
. a.' t e s.
.c d ' ' * '_-, -.'_"._~....i'.*'..a.=
w'_o s"-a.
a...' cd_- - ' e.' a.e'.
a
.7 f.
yc se; at > CC'/c cr. en will anticit. ate closure of the turhine ste valves.
"he F_ T is net given credit for any other initiacing evenes.
acn. n system may be =anually.ev..cassed b.v use of a.e.vs._.t e n (<,'cuc-x c
, 4. e..,., e.4_..
9_..
C.6,., 7.e.
..w.w. _a..a1_ 1.
Eeeee
.r 4.
.n\\f b.wa_.an'
- 4. e.
24..
4 S'
( <,20.
- y.. T, a-a
_s_.... ' a. a d...-.-...c."_ca_y
.e. a_.. C a.
'_..3 :,,a.e =.
-.e. 4_.. A. 4..
9...
.. w. A.
,. 9.....
aJ
_4
- 6. w..
....g
.c_9.e r.
\\' J.. e,
A.g e o.m..
- .J_
....s.e.
e ey.et e...,
A v O_:?...g_.
J.e
. e. e.
..J.
- y. " J 'v..
.w
.c..
g.
wo.
_J..A e..a.
- --a..e.
_1
- e..e.,
- a..#.7. a...-=_' a Co. \\.
'.k..
.e...e a. -. a_c-.e
-.#..*9. - -
~
A
.y
..J
.g
.c y
.w yy.
c...
S-a s*o-'.a *. '_o's
^#
y "a*
, ku'. ' '..*. ~..' 07 '_'".3'_".
- 4..m
. >4
. 4.ed
.vi
,.....,.,.r, %.. s e
_4.sti.
.e e
.i.9_ W C _
_a 4 3..c.,..,
._'....J..
~...
4 e.
2 w.
y ws
.... y s,. s4 3..,
- w.
9.n 3_a.
4.
.. 4 4
..... a.. d A.o +r.... e.,
6.k... +,.4.-
C._4 9.. ;-,. + w e.e
...y w.
p...
~#
7v 4" v"' '.P,4 e..
.k Q +- ' ** #
.e. s..e e 4*
e.e e
a.
- e..v.4 - e e. J_
g g.i.. **g '.,.
C.'m4 m
w
_y y..
4" C "., - A C
.# a. *.'. '. b.16*.'_.
e " e..... e 7 -.=. e..e '. n.
.s.
y an. J. 4..
' b...
w. A.,, A..,.4. e.
a'. e
- = =.
A e.
y.
.c
..e
.,4.e 7_
3 e.. ~
c.
p.....
._4.w. 4..
.e J_. ag e. +,.,.,.e e.
e e 4e.
e...a_ a.. a., -
f.
'.' c.'. *. ~. S '. _ a.
.a...
4...
r...._.
- a. w...e. _4 3 3.re
~
. C c~....'"j'..a".'.
, y c.
c 3..
.y 4,, + w.., g.er. g, +>.
e... e,. a.. c c. v.e.,
e
.o nn._1. w. y e. 4 n.
. e,.u. 1. 2..
e.. e. g. 4 -
w..c -
g.
m..
..v..
n-c
.g..
'.'.e cd. * *.t.'.'
". e c-k.e. *. ',,o
'_y,
'.".a. -, e.* a.o.
4_ s ". a 's _4. e d. '.-..= _.".a 7 9 -. a..=. a..
y the breaker.
To.;be effective the F.FT must te initiated virtually i=sediate'_y.
T7A has
.-.a' g e.,d_ s e.'n we
'.k..-'.,...a..a1 4_.d ','_ E'.d C..
....~.s'.
- e. a. ec 'w.4a'..'..e4.
b c
c c
.a
..a
.,,. p.a_
c.t.e.,.
w...e u.
.e u_
4
-,,_ e
.4_., c., + w..
.,. r e,.,, a +o..,,
4 f...
- c. _....
.m..
when autccatic acticn should have cecu red will not have a sig i'icant i rovement en the situatien.
-'he pcwer to the =ctor-generater sets can be I,'
g tripped manually frcm the control rec =.
Therefere, previsiens fcr ma.ual initiation cf the 500 FST feature are unnecessary.
. As described above, a fast closure sensor from each of two turbine control valves provides input to one RPT system; sensors from the other two turbine control valves provide inputs to the second RPT system.
Similarly, position switches for each of two turbine stop valves provide input to one RPT system; sensors from the other two stop valves provide inputs to the other RPT system. The staff performed a failure mode analysis of the system. The most severe postulated failure would be the failure of one of the' four turbine control valves or turbine stop valves to close (thus not providing an actuation signal to the one RPT division) with simultaneous failure of the bypass valves to open and simultaneous failure of the other RPT system. However, in this case,
the open line to the turbine would in effect be acting as a byoass.
The licensee was requested to verify that the RPT feature need not actuate for the condition where one of either the turbine stop or control valves does not close.
The licensee referenced calculation of load rejection with bypass and turbine trip with bypass as supportive evidence. We have reviewed the flow capacities of the bypass and open steamline configurations and have evaluated the dynamic pressure response associated with ea:h configuration.
We have concluded that the pressurization transient analyses with bypass provide a conservative estimate of the effect of failure of one or more turbine stop or control valve to close.
The results of the pressurization transient with bypass and vithout RPT indicate that the transient is not limiting.
Therefore, the design of the RPT system to respond for only the condition where all turbine stop or control valves close is acceptable.
3.4 RPT Testability Capability to check the RPT sensors and logic is provided by operating each valve one at a time.
Lights across the relay contacts in the logic indicate proper operation at that point.
The RPT systems do not need to bt. bypassed to conduct such tests. However, during the periodic testing of the scram logic, where two valves are operated simultaneously, the affected RPT system must be bypassed briefly to prevent RPT actions.
Appropriate technical specifications cover this situation. The RPT circuit breakers will be functionally tested during refueling outages.
. The ??? feature is recnired to interrupt the pu=p motor circuit V. thin 175 =i111 seconds of the start of valve closure.
Of this,10 =s ' ; alletted for syste= acticn and senser response, 30 ms for logic restense, and 135 :s for breaker action. Because the IOC ?ST feature = st function so prc=ptly, we require that appropriate response time testing be conducted at each refueling shutdown. T7A is not in ec=plete agree =ent with this staff requirement, but has com=itted (in a letter dated Janna:y 23,197?,
J. I. Gilleland to T. A. Ippolito) to submit their proposal en this matter no later than July 23, 1C70 (well before the next refueling). The ??I system V" be testei preeperational te ver fy response time.
The final design of the RPT feature for Browns Ferry Unit No.1 adequately complies with IEEE Standard 279 for its stated purpose. All parts of the RPT feature are appropriately qualified to mitigate appropriate anticipated pressure transients from the turbine. We conclude, therefore, that the design is acceptable. The matter of periodic (refueling interval) response time testing is being deferred and must be completed prior to the next refueling outage for this unit.
4.0 Environmental Considerations 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. fiaving made this determination, we have further concluded that the amendment involves an action ~which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
5.0 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 reason-i able assuiance that the health and safety of the public will not be endangered by operation ir the proposed manner, and (3) such activities will be conducted in compliance with the Commisrion's regulations and the issuance of this amendment will not be inimical to the
, mon defense and security or to the health and safety of the public.
Dated: February 8, 1979
References:
1.
Letter, TVA (Gilleland) to USNRC (Ippolito) dated January 23, 1979.
2.
" Generic Relo.ad FJe1 Application," General Electric Report, NEDE-240ll-P-3, Jated March 1978.
3.
USNRC letter (Eisenhut) to General Electric (Gridley) dated May 12, 1978, transmitting " Safety Evaluation for the General Electric Topical Report, ' Generic Reload Fuel Application,' ( N EDE-240l l -P ). "
4 TVA letter (Cilleland) to USNRC (Denton) dated November 30, 1978,
" Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Plant Unit 1 Reload 2,"
NED0-24136, Rev.1, November 1978.
5.
Carmich'ael, L. A.,' and Niemi, R. 0., " Transient and Stability Tests at Peach Bottem Atomic Power Station Unit 2 at End of Cycle 2,"
EPRI-NP-564, June 1978.
6.
Letter, R. Engel (GE) to Office of Nuclear Reactor Regulation (NRC),
dated July 11, 1977.
7.
Letter, E. D. Fuller (GE) to U. S. Nuclea
- egulatory Commission, dated October 25, 1977.
8.
" Impact of One-Dimensional Transient Model on Plant Operating Limits,"
enclosure of letter, E. D. Fuller (GE) to V. S. Nuclear Regulatory Commission, dated June 26, 1978.
9.
Memo from P. S. Check (RSB-NRC) to T. A. Ippolito (ORB #3-NRC),
" Browns Ferry 3 - Cycle 2 Reload (TACS #8026)," November 8,1978.
10.
NRC letter (Ippolito) to TVA (Hughes), Amendment Nos. 45, 41, and 18 to Facility License No. DPR-33, DPR-52, and DPR-68 for Browns Ferry Nuclear Plant Units Nos.1, 2, and 3, dated November 18, 1978.
- 12. Safety Evaluation Report related to operation of Edwin I. Hatch Nuclear Plant, Unit No. 2, Georgia Power Company', et. al., USNRC, Office of Nuclear Reactor Regulation, Docket No. 50-366, NUREG-0411, June 1978.
i3. Letter, TVA (Gray) to USNRC (Case) dated July 18, 1978, TVA BFNP TS 111.