ML20151K819
| ML20151K819 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 05/08/1985 |
| From: | Rubenstein L Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20151K803 | List: |
| References | |
| FOIA-85-511 NUDOCS 8505160436 | |
| Download: ML20151K819 (6) | |
Text
_
)E f s1 J
,/
),
UNITED STATES j
NUCLEAR REGULATORY COMMISSION y
e a
wasMNGTON,0. C. 206S5
%,+..../
MAY 8 1985 MEMORANDUM FOR:
T. M. Novak, Assistant Director for Licensing, OL FROM:
L. S. Rubenstein Assistant Director for Core and Plant Systems, DSI
SUBJECT:
TECHNICAL SPECIFICATION FOR RIVER BEND The Core Performance Branch has reviewed the Final Draft copy of the proposed Technical Specifications for River Bend Station. The list of Specifications reviewed is given in Enclosure 1. contains our comments on the proposed Specifications.
.N
(
u <_
L. S~.
ubenstein, Assistant Director for C e and Plant Systems, OSI
Enclosures:
As stated cc:
R. Berneru H. Thortpson A. Schwencer
.S. Stern D. Houston
Contact:
W. Brooks, CPB:DSI S. Sun, CPB:DSI X-29498 X-29499 s
QYoll(oOY%Xh x
/,
s~
g
4 List of Specifications Reviewed - River Bend Definitions Specifications 1.2 2.1.2 1.3 2.2.1(IRMandAPRM) 1.7 3/4.1.1
~
1.8 3/4.1.2 1.9 3/4.1.3.1(except4.1.3.1.4)
I 1.15 3/4.1.3.2 1.16 3/4.1.3.4 1.21 3/4.1.3.5 1.22 3/4.1.3.6 1.25 3/4.1.4.1 3
1.29 3/4.1.4.2 1.34 3/4.2.1
, l 1.37 3/4.2.2 i
1.40 3/4.2.3 1.45 3/4.2.4 Table 3.3.1-1 (IRM and APRM only)
Table 3.3.1-2 (IRM and APRM only) 3/4.3.6 (except Table 4.3.6-1) 3/4.3.7.6 3/4.3.7.7 3/4.3.7.9 3/4.4.1.2 3/4.4.1.1 3/4.4.1.2 3/4.9.1 3/4.9.2 3/4.9.10.1 3/4.9.10.2 3/4.10.2 3/4.10.3
~
1 5.3 5.6
Comments on Technical Specifications - River Bend Specification 3/4.2.1 and Figure 3.2.3-2 There are MAPLHGR curves for five different fuel enrichments given in this Specification. However, Chapter 4.3 of the FSAR lists only 3 enrichments in the core. It is our understanding that a revision to the FSAR is being prepared to correct the confusion.
It should be noted that the i 1eric Rod Withdrawal Error analysis which is the basis.of Technical Specif i ation Figure 3.2.3-2 is not valid for the five enrichment core loading. A compliance check must be provided for this analysis as part of the FSAR revision.
Specification 3/4.4.1.1 This specification as written does not address the issue of monitoring core stability in accordance with the recommendations of GE letter SIL-380. We require such monitoring and the inclusion of the requirement in the Technical Specif'ications. An acceptable Specification is attached. It is our under-standing that the appifcant is currently performing the analysis to support the inclusion of the revised Specification.
e
.o.
s 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM pgQ][ { ((] Gy RECIRCULATION LOO 75 LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation with:
Total core flow greater than or equal to 45% of rated core flow, or a.
THERMAL POWER less than or equal to the limit specified in Figure b.
3.4.1.1-1.
APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.
ACTION:
With one reactor ecolant system recirculation loop not in operation, a.
immediately initiate action to reduce THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1-1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and initiate measures to place-the unit in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
With no reactor coolant system recirculation loops in operation, immediately initiate action to reduce THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1-1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and initiate reasures to place the unit in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
.With two reactor coolant system recirculation locps in operation and c.
total core flow less than 45% of rated core flow and THERMAL POWER greater than the limit specified in Figure 3.4.1.1-1:
1.
Determine the APRM and LPRMa* noise levels (Surveillance 4.4.1.1.2):
a)
At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and b)
Within 30 minutes after the completion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER.
l 2.
With the APRM or LPRh** neutron flux noise levels greater than three times their established baseline noise levels, immediately initiate corrective action to restore the noise levels to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> by increasing core flow to greater than 45% of rated core flow or by reducing THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1-1.
~
"See Special Test Exception 3.10.4.
t
APRoaless l
PERRY - UNIT 1 3/4 4-1.
L
tb bLi CDR REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.1.1 Each reactor coolint system recirculation loop flow control valve shall be demonstrated OPERABLE at least once per 18 months by:
Verifying that the control valve fails "as is" on Ipss of hydraulic a.
pressure at the hydraulic control unit, and b.
Verifying that the average rate of control valve movement is:
1.
Less than or equal to 11% of stroke per secon'd opening and 2.
Less than or equal to 11% of stroke per second closing.
4.4.1.1.2 Establish a baseline APRM and LPRM" neutron flux noise value within the regions for which monitoring is required (Specification 3.4.1.1, ACTION c) within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of entering the region for which monitoring is required unless baselining has previously been performed in the region since the last refueling outage.
- Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.
i Apa ct W PERRY - Ut4IT 1 3/4 4-2
l 8
m
- =
1
\\
so a
3 E
c:
a' i
2:
r-af
-4 a
N g
50
~
I
-m 4
e
...na,,r 8,mP a
i_
~
a
- t ac I
W
-.n*8 40 g
O d.'
2
+
w I
e 30 s
1
,;j m
t F
1 e
i w
20
.6
_3_--
t to ----
i L,
c.3 T
1 O
l
-m gr.
30 40 50 80 70 7
D l g CORE FLOW (% R ATED)
~!.
c.
TilERHAL POWER VERSUS CORE FLOW I
am W
t FIGURE 3.4.1.1-1 o
D e
1 b
Q9 1