ML19253B835
| ML19253B835 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/20/1979 |
| From: | Herbein J METROPOLITAN EDISON CO. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| 780420-01, 780420-1, GQL-0748, GQL-748, NUDOCS 7910310589 | |
| Download: ML19253B835 (8) | |
Text
r)64f21'?Y
~
k REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)50-289 DISTRIBUTION FOR INCOMING MATERIAL REC: REID R W ORG: HERCEIN J G DOCDATE: 04/20/7S DATE RCVD: 04/26/78 METROPOL EDISON NRC COPIES RECEIVED DOCTYPE: LETTER NOTA.7 I ZED: NO LTR 1 ENCL 1
SUBJECT:
INFO TO APPLICANT"S PROPOSED TECH SPEC CHANGE REQUEST NO 70A, FURNISHING ADDL THAT WERE INADVERTENTLY CONSISTING OF TECH SPEC PAGES SPECIFIC TO CYCLE S,AND FURNISHING JUSTIFICATION FOR OMITTED FROM APPLICANT"3 CYCLE 4 SUBMITTAL, THESE CHANGES W/ATT I REVIEWER INITI AL:
XM PLANT NAME: THREE MILE ISLAND - UNIT 1 DISTRIBUTOR INITIAL: D
- DISTRIBUTION OF THIS MATERIAL IS AS FOLLOWS ******************
GENERAL DISTRIBUTION FOR AFTER ISCUANCE OF OPERATING LICENSE.
(DISTRIBUTION CODE AOO1)
FOR ACTION:
BR CHIEF REID**W/7 ENCL INTERNAL:
LE**W/EN NRC PDR**W/ ENCL OELD**LTR ONLY av a we. crmu HANAUER**W/ ENCL CHECK +*W/ ENCL EISENHUT**W/ ENCL SHAO**W/ ENCL BAER**W/ ENCL BUTLER **W/ ENCL EEB**W/ ENCL J COLLINS **W/ ENCL J.
MCGOUGH**W/ ENCL EXTERNAL:
LPDR'S HARRISBUFG, PA**W/ ENCL TIC **W/ ENCL NSIC**W/ ENCL ACRS CAT B**W/16 ENCL 1505 151 DISTRIBUTION:
LTR 40 ENCL 39 CONTROL NDR:
781160186 SIZE: 2P+5P THE END
- +w *****+<
/
w 7910310 6 8 9
/
/
Q:f70
,tc.m. -
/;,
M44 900 #tOGef33 METROPOLITAN EDISON COMPANY POST OFF!CE Box 542 RE ADING. PENNSYLVANI A 19603 TELEPHONE 215 - 929-2601 1
April 20, 1978 G3 GQL 0748 EP,
-3 Director of Nuclear Reactor Regulation C....5 1~
Attn:
R. W. Reid, Chief
- .y Operating Reactors Lranch No.
- 9 U. S. Nuclear Regulatory Comrs.., ion G
98 x
^~
Washington, D.C.
20555
Dear Str:
Three Mile Island Nuclear Station, Unit 1 (TMI-1)
Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 70A (Additional Information)
Your Mr. G. B. Zwetzig identified several TMI-1 Technical Specification pages, specific to Cycle 3, that were inadvertently omitted from our Cycle 4 submittal. These pages and justification for their changes are as follows:
.; E FOE'[
r(n-c -[ {CU Lfig,,1 j v, ;/.1 ; 1.LL U Ui a Figure 2.1-3 The power 12vels for Curves 2 and 3 of Figure 2.1-3 of the TMI-1 Cycle 4 Technical Specifications should be consistent. These power levels differ from those reported for Cycle 3 by the setpoint adjustment error which is a necessary component of the calibration and instrumentation errors.
The correct Cycle 4 values for the power levels of the aforementioned Curves 2 and 3 are 87.1% and 59.6% respectively. Figure 2.1-3 has been revised to reflect these valves. There is no reduction in the safety margin as a result of these changes.
Fizure 2.1-1 Section 6 of the Cycle 4 Reload Report submitted in support of Tech. Spec.
Change Request No. 70, states that the only dif ference between cycles 3 and 4 is the core configuration. For Cycle 4, the addition of the higSer flow resistance Mark B2 and B3 assemblies provides additional DNBR margin over Cycle 3 due to increased flow through the limiting Mark B4 assembly.
Additional conservatism for Cycle 4 operation is provided by the reduced peaking factors. The minimum DNBR calculated from the Cycle 2 analysis (the Cycle 2 DNBR analysis was used for reference Cycle 3) is based 1505 152 s
d,l 1
April 20, 1978 R..W. Reid GQL 0748 on a radial-local peak of 1.783.
The maximum radial-local peak calculated for Cycle 4 operation (287.1 EFPD Cycle 3 and Non-cross core shuffle), in-cluding 8% nuclear uncertainty, is 1.547 at BOC.
This decreases to 1.403 at the end of Cycle 4.
This provides a 13.2% margin to the design peak at POC and a 21.3% margin at E0C.
Figure 2.1-1, unchanged for Cycle 4, represents a more conservative limit and does not result in a reduced safety margin.
Page 2-4, Page 2-9 (Table 2.3-1), and Figure 2.3-1 A bounding analysis, showing peak RC system pressure vs. moderator coefficients and a conservative doppler coefficient for the feedwater line break accident, has been performed to justify the high pressure trip setpoint of 2405 psig and the corresponding 2500 psig pressurizer code safety valve settings.
Re-suits of this analysis were submitted to NRC on April 17, 1978, (GQL 0669),
and supplemented on April 20, 1978 (GQL 0743). Since the Cycle 4 parameters are bounded by the analysis, and the 2750 psig reak pressure limit is not exceeded, there is no reduction in safety margin and no resulting threat to the health and safety of the public.
erely,/
Si
\\ ^/
}w it
/
. G. Herbein Vice President - Generation b
JGH:RJS;jmr Attacht ents
)
2400
\\
?
2200 E
[
E 2
5 0
E 2000
/
/
/
3 3
f E
/
o 1800 t
1600 550 580 B00 620 640 660 Reactor Outiet Temperature, 'F REACTOR COOLANT FLOW CURVE (LBS/HR)
POWER PUMPS OPERATING (TYPE OF LIMIT) 1 139.2 x 103 (100r,)*
112%
Four Pumps (DNBR Limit) 2 104.5 x 106 (74.7",)
87.1%
Tnree Pumps (DNSR Limit) 3 68.8 x 105 (49.2%)
59.6%
One Pump in Eacn Loop (Quality Limit)
- 105.5% of Cycle 1 Design Flew THREE MILE ISLAtlD, UNIT I l
ORE PROTECTION SAFETY BAbis Figure 2.1-3 1505 154
2600 2400 ACCEPTABLE
{
SPERATION 2203 52 E:
5 2000 e
/
8 f
UNACCEPTABLE OPERATIO:i 1800
/
i 1600 560 380 600 620 640 660 Reactor Outlet Temperature, F THREE MILE ISLAND, UNIT I COR: PROTECTION SAFETY LIMIT Figure 2.I-I 1505 155
2.2 SAFE'"Y LIMITS - REACTCR SYSTD! PRESSUPS Applicability Applies to the limit en reactor coolant system pressure.
Obj ective Tc maintain the integrity of the reacter coolant system and to prevent the release of significant amounts of fission prcduct activity.
_ Spec ification 2.2.1 The reactor coolant system pressure shall nct exceed 2750 psig when there are fuel assemblies in the reactor vessel.
Bases The reactor coolant system (1) serves as a barrier to prevent radionuclides in the reactor coolant from reaching the at=csphere.
In the event of a fuel cladding failure, the reactor coolant system is a barrier against the release of fission products. Establishing a system pressure limit helps to assure the integrity of the reactor coolant system.
The maximum transient pressure allevable in the reactor coolant system pressure vessel under the ASME Code,Section III, is 110% of design pressure. Thus, the safety limit of 2750 psig (110% of the 2500 psig design pressure) has been established.
(2) The maxi =um settings for the reactor high pressure trip (2h05 psig) and the pressurizer code safety valves (2500 psig) (3) have been established in accordance with ASME Boiler and Pressure Vessel Code,Section III, Article 9, Winter,1968 to assure that the reactor coolant system pressure safety limit is not exceeded. The initial hydrestatic test was corducted at 3125 psig (125% of design pressure) to verify the integrity of the reactor coolant system. Additional assurance that the reactor coolant system pressure dces not exceed the safety limit is provided by setting the pressuriser electromatic relief valve at 2255 psig. (h)
References (1) FSAR, Secticn h (2) FSAR, Section h.3.10.1 (3) FSAR, Section 4.2.h (h) FSAR, Table h1 1505 156 2_3
TABLE 2.3-1 RFACTOR PROTECTION SYSTEM TRIP SETTING LIMITS Four Reactor Coolant Three Reactor Coolant One Reactor Coolant Pumps Operating Pumps Operating Pwnp Operating in (Hominal Operating (Nominal Operating Each Loop (Nominal Shu+ down Power - 100%)
Power - 75%)
Operating Power h9%)
Ilypass 1.
Nuclear power, Max.
105.5 105.5 105 5 5.O(3)
% of rated power 2.
Nuclear Power based on 1.08 times flow minus 1.03 times flow minus 108 times flow minus Bypassed h
flow (2) and imbalance reduction due to reduction due to reduction due to max. of rated power imbalance (s) imbalance (s) imbalance (s) 3.
Nuclear power based NA NA 91%
Bypassed (5) on pump monitors, max. % of rated power h.
High reactor coolant 2h05 2h05 2h05 1720(
system pressure, psig max.
4 5
Low reactor coolant 1800 1800 1800 Bypassed system pressure, psig min.
6.
Variable low reactor (11.75 Tout-5103)
(11.75 Tout-5103)(1}
(1175 Tout-5103)(
Hypassed 1
coolant system pressure psig, min.
LJ1
'(.
Reactor coolant 1-...g.
619 o
619 619 619 F., max.
W 8.
liigh Reactor Building h
h h
h pressure, psig, max.
LJ1 N
(1) Tout is in degrees Fahrenheit (F)
(2) Reactor Coolant System Flow, %
(3) Administratively controlled reduction set only during reactor shutdown (h) Automatically set when other segments of the RPS (as specified) are bypassed (5) The pump monitors also produce a trip on:
(a) loss of two reactol coolant pumps in one reactor coolant loop, arrl (b) loss of one or two reactor coolant pumps during two-pwup operation (6) Trip settings limits are setting limits on the setpoint side of the protection system bistable comparators.
I
2500 P= 2405 psig 2300 y
ACCEPTABLE ay e
OPERATION c.
f 2100 P
o I
0 40 E
1900 4
UNACCEPTABLE P = 1800 psig OPERATION E
E 1700 1500 540 560 580 600 620 640 Reactor Outlet Temperature, F Three Mile Island, Unit 1 l
Protection System Maximum Allowable Set Points Figure 2.3-1 1505 158