ML20236X121
| ML20236X121 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 10/26/1987 |
| From: | Calvo J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20236X123 | List: |
| References | |
| NUDOCS 8712080436 | |
| Download: ML20236X121 (10) | |
Text
'
f y,
~
UNITED STATES -
.:E
'g' NUCLEAR REGULATORY COMMISSION e
5
- ' l-
' W ASHINGTON, D. C. 20555
%....../
GULF STATES UTILITIES COMPANY l-
- DOCKET NO. 50-458 RIVER BEND STATION, UNIT 1 i
AMENDMENT-TO FACILITY OPERATING LICENSE--
Amendment No.13.
License No. NPF-47
.1.
The Nuclear Regulatory. Commission (the Commission or the' NRC) has found that:
A.
The application'for amendment filed by Gulf. States. Utilities Company, dated August 14, 1987, complies with the studards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's regulations set forth in 10 CFR Chapter I; B '.
.The facility will operate in conformity with the application, the-provisions of the Act, and the. regulations of the Commission; C.
There is' reasonable assurance:. (i) that the activities authorized by this amendment can be conducted without endangering the health ~
and safety of the public, and '(ii) that such activities will be -
conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not.be inimical to the common defenseLand security or.to the health and safety of the public;.and l
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all. applicable requirements have been satisfied.
AY$
2.
.Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment and l
paragraph 2.C.(2) of Facility Operating License No. NPF-47 is hereby l
amended to-read as follows:
(2) Technical Specifications and Environmental Protection Plan
' The. Technical { specifications contained in Appendix A, as revised through and the Environmental Protection Plan contained in' Appendix B, Amendment No.
3 are hereby incorporated in the license.
GSU shall operate the-facility in accordance with the Technical Specifications and the Environmental Protection Plan.
1 1
l f
h P
j
.P a
(m-q y K.,,,
- r.
^
x.
~
c' .
. p;
\\,
f
~
- -- 2 :-
f,,
y l
[
l 3.
-This'licen'se am'endment'is effective'as of its date,oflissuance.
- h.,
FOR THE NUCLEAR-REGULATORY COMMISSION 1
3
> x Ri%
m Jose A. Calvo, Director.
- 4,-
Project'. Directorate'- IV 1
Division of' Reactor Projects'-_III,
.IV,.V'and Specia1 Projects Office.of. Nuclear Reactor' Regulation
Attachment:
, Change's' to the' Technical.
4 Specifications
' October 26,-1987 Date of Issuance:
H<
[.
i l
l u
l' L
L_----
?Q,
uc
~
~~
-~ ~ T '---
~ - - - ~ - ~ ' n 4
r m
3 1
y
..: n
-4 N'
4,
(
' ATTACHMENT TO LICE'NSE AMENDMENT NO~13 '
L.
'.. FACILITY'0PERATING LICENSE NO. NPF-47
~
DOCKET NO.-50-458L R'eplace'.the following' page o'f the' Appendix "A'.'STechnical Specifications' with
. a p:
- the. enclosed'page. 'The" revised.page is identified by' Amendment number and J.l,,
contains.a: vertical line-indicating the. area of' change. Overleaf. page provided to maintain document completeness.
' REMOVE INSERT
,m, xxii
' xxii 3/4 1-19-.
3/4 1-19 3/4 1-20 3/4/ 1-20
-3/4'1-21 3/4 1-22 G.a B 3/4 1-4 B 3/4'1 h B 3/4 1 B 3/4 1-5 I
L 0
I:
1 i
)
di __ __ __1._ __ _. _ _ _ _
__m____
i
w
.+a hp w-
.n
- f. r '
e
.;n INDEX ADMINISTRATIVE CONTROLS--
- SECTION-P PAGE 6.13 PROCESS CONTROL PROGRAM-(PCP)................................
6 6.14: 0FFSITE DOSE ~ CALCULATION' MANUAL-(00CM).......................-
6-23
' 6.15-MAJOR'CH'ANGES TO'RADI0 ACTIVE LIQUID, GASE0US' AND SOLID WASTE-i.
TREATMENT SYSTEMS................................
.6-23 I
'kr i
oj
.j
)
i-
..gQ; l
'l l
l 1
1 i
r u-RIVER. BEND - UNIT 1 xxi
INDEX LIST OF FIGURES i
FIGURE TITLE PAGE l
3.2.1-1 Maximum Average Planar Linear Heat Generation "l
Rate (BP85RB094)....................................
3/4 2-2 3.2.1-2 Maximum Average Planar Linear Heat Generation Rate (BP85RB163)....................................
3/4 2-3 3.2.1-3 Maximum Average' Planar Linear Heat Generation Rate (BP8 SRB 248).....................................
3/4 2-4 3.2.1-4 Maximum Average Planar Linear Heat Generation l
Rate (BP8 SRB 278).....................................
3/4 2-5 i
3.2.1-5 Maximum Average Planar Linear Heat Generation Rate (BP8 SRB 299).....................................
3/4 2-6 3.2.1-6 Maximum Average Planar Linear Heat Generation Rate (BP85RB305).....................................
.3/4 2-6A 3.2.3-1 MCPR..............................................
3/4 2-9 f
3.2.3-2 MCPR..............................................
3/4 2-10 p
3.4.1.1-1 Thermal Power versus Core Flow.....................
3/4 4-3 3.4.6.-1-1 Minimum Temperature Required Versus Reactor Pressure...........................................
3/4 4-24 4.7.4-1 Sample Plan for Snubber Functional Test............
3/4.7-15 B 3/4 2.3-1 Power Flow Operating Map...........................
B 3/4 2-6 B 3/4 3-1 Reactor Vessel Water Leve1.........................
B 3/4 3-8 B 3/4.4.6-1 Fast Neutron Fluence (E>l MeV) at 1/4 T as a Function of Service Life...........................
B 3/4 4-8 1
5.1.1-1 ExclusionArea93g.................................
5-2 5.1.2-1 Low Population Zone................................
5-3 5.1.3-1 Map Defining Unrestricted Areas and Site Boundary for Radioactive Gaseous and Liquid Effluents.......
5-4 6.2.1-1 RBNG Organization..................................
6-3 6.2.2-1 River Bend Station Organization....................
6-4 1
l 1
l RIVER BEND - UNIT 1 xxii Amendment No. 13 1
.. E i JRE' ACTIVITY CONTROL: SYSTEMS J3/4.1.5' STANDBY LIQUID CONTROL SYSTEM LIMITING-CONDITION FOR OPERATION'
.3'.1.5 Two standby liquid control' subsystems shall-be OPERABLE.
APPLICABILITY: ' OPERATIONAL CONDITIONS 1, 2 and 5*.
f ACTION:
a.
In OPERATIONAL CONDITION 1 or 2:
1.
With one subsystem inoperable, restore'the inoperable subsystem.
to OPERABLE status within 7 days or'be'in at least HOT' SHUT 00WN-within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2.
With both subsystems inoperable, restore at least one subsystem
~
to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be 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.
In OPERATIONAL CONDITION 5*:
1.
With one subsystem inoperable, restore the inoperable subsystem-
.to'0PERABLE status within 30. days or insert.all insertable con-trol rods within the.next hour.
2.
With both subsystems inoperable, insert all insertable control-rods within one hour.
SURVEILLANCE'E-REQUIREMENTS.
4.1.5 Each standby liquid control ~ subsystem shall be demonstrated OPERABLE:
a.
At-least.once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that; 1.
The temperature sodium pentaborate solution is greater than or equal to 45'F 2.
The available volume of sodium pentaborate solution is greater than.or equal to the minimum required available volume deter-mined once per 31 days per Specification 4.1.5.b.2.
b.
At least once per 31 days by; 1.
Verifying the continuity of the explosive charge.
- With any control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
l 1,
L RIVER BEND - UNIT 1 3/4 1-19 Amendment No. 13
=
p o,
i
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.
Determining *, that the available weight of Boron-10 is greater than or equal to 143 lbs, the percent weight concentration of sodium pentaborate in solution is equal to or less than 9.5%
by weight, and the minimum required solution volume.
3.
Verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
4.
Determining that the Standby Liquid Control System satisfies the following equation:
(C)(E) > 413 Where:
C = sodium pentaborate concentration, in weight percent, as determined per specification 4.1.5.b.2.
E = Boron-10 enrichment, in atom percent **.
c.
Demonstrating that, when tested pursuant to Specification 4.0.5, the minimum flow requirement of 41.2 gpm per pump at a pressure of greater than or equal to.1220 psig is met, d.
At least once per 18 months during shutdown by; 1.
Initiating one of the standby liquid control system loops, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel.
The replacement charge for the explosive valve shall be from the same manufactured batrAias the one fired or from another batch which has been certified by having one of that batch success-fully fired.
Both injection loops shall be tested in 36 months.
i l
l l
- This test shall also be performed anytime water or boron is added to the solution or when the solution temperature drops below 45 F.
l
- The Boron-10 enrichment of the solution shall be determined anytime boron is added to the solution.
RIVER BEND - UNIT 1 3/4 1-20 Amendment No. 13
l' L
BASES CONTROL RODS (Continued)
. Control rod coup 14ng integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position feature provides the only positive means of determining that a rod is properly coupled and, therefore, this check must be performed prior to achieving criticality after completing CORE ALTERATIONS that could have affected the control rod coupling integrity.
The subsequent check is performed as a backup to the initial demonstration.
In order to ensure that the control rod patterns can be followed and, therefore, that other parameters are within their limits, the control rod position indication system must be OPERABLE.
The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure.
The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system.
The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.
The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
3/4.1.4 ROD PATTERN CONTROL SYSTEM The rod withdrawal limiter system input power signal originates from the
]
first stage turbine pressure.
When operating with the steam bypass valves open, this signal indicates a core power level which is less than the true
{
core power.
Consequently, near the low p3er setpoint and high power i
setpoint of the rod pattern control syster,' the potential exists for non-conservative control rod withdrawals.
Therefore, when operating at a sufficiently high power level, there is a small probability of violating fuel Safety Limits during a licensing-basis rod withdrawal error transient.
To ensure that fuel Safety Limits are not violated, this specification prohibits control rod withdrawal when a biased power signal exists and core power exceeds the specified level.
Control rod withdrawal and insertion sequences are established to assure that the maximum in-sequence individual control rod or control rod segments which 1
are withdrawn at any time during the fuel cycle could not be worth enough to l
result'in a peak fuel enthalpy greater than 280 cal /gm in the event of a control I
rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater i
than 20% of RATED THERMAL POWER, there is no possible rod worth whic.h, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm.
Therefore, requiring the RPCS to be OPERABLE, when THERMAL POWER is less than or equal to 20% of RATED THERMAL POWER, provides adequate control.
RIVER BEND - UNIT 1 B 3/4 1-3
R_EACTOR COOLANT SYSTEMS BASES I
R0D PATTERN CONTROL SYSTEM (Continued)
The RPCS provides automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.
The analysis of the rod drop accident is presented in Section 15.4.9 of the FSAR and the techniques of the analysis are presented in a topical report (3) and two supplements (2, 3),
The RPCS is also designed to automatically prevent fuel damage, during higher power operation, in the event of erroneous rod withdrawal from locations of high power density.
A dual channel system is provided that, above the low power setpoint, restricts the withdrawal distances of all non peripheral control rods.
This restriction is greatest at highest power levels.
3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern.
To meet this objective it is necessary to inject a quantity of Boron-10 which produces a concentration of 122 ppm in the reactor core and other piping systems conne.cted to the reactor vessel.
This concentration is increased by 25% to allow for potential leakage and imperfect mixing.
The required concentration is achieved by having a minimum available quantity of 143 pounds of Boron-10 contained in l
the net amount of solution which is above the pump suction, thus allowing for theportionthatcannotbeinjected.
The pumping rate of 41.2 gallons per minute (gpm) per pump provides a negative reactivity insertion rate, which ade-l quately compensates for the positive reactivity effects due to temperature and xenon decay during shutdown.
~
g The Standb Liquid Control System is also required to meet the criteria of 10CFR50.62, y' Requirements for reduction of risk from Anticipated Transients Without Scram (ATWS) Events For Light-Water-Cooled Nuclear Power Plants" by having the equivalent control capacity of a 66 gpm system using 13 weight per-cent natural sodium pentaborate.
The equivalency requirement is fulfilled by having a system which satisfies the equation given in surveillance require-ment 4.1.5.b.4.
Each parameter is tested at an interval consistent with the 1.
C. J. Paone, R. C. Si t rr, and J. A. Woolley, " Rod Drop Accident Analysis for Large BWR's," G. E. lepical Report NED0-10527, March 1972 2.
C. J. Paone, R. C. Stirn and R. M. Young, Supplement 1 to NED0-10527, July 1972 3.
J. M. Haun, C. J. Paone and R. C. Stirn, Addendum 2 " Exposed Cores,"
Supplement 2 to NE00-10527, January 1973 l
RIVER BEND - UNIT 1 B 3/4 1-4 Amendment No. 13 l
l
Q_~~
REACTOR C00LANT' SYSTEMS BASES-
-STANDBY LIQUID CONTROL SYSTEM (Continued) potentia 1L for that' parameter to vary and also to assure proper equipment per-formance where: applicable.
Enrichment testing is only: required when boron addition occurs since change cannot occur by any other process.
. ith redundant pumps and' explosive injection valves and with a highly W
reliable control rod scram system, operation of the reactor'is permitted to
' continue for= short periods of time with the system aoperable or for. longer V.
periods 'of time with' one of the redundant components' inoperable.
Surveillance requirements are established on a frequency that assures a high reliability of the. system. Once the solution is established, Boron-10 l
concentration will not vary unless more boron or water is added.
Therefore, a check on the' temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solu-tion is available for use.
Replacement of the explosive charges in the valves at regular intervals will assure that these ' valves will not fail due to deterioration of the chargesi L
RIVER BEND
. UNIT 1 B 3/4 1-5 Amendment No. 13 '
L
_ _ _ _ _ _ _ _