ML19344B168
| ML19344B168 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 08/19/1980 |
| From: | Grace R VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | |
| Shared Package | |
| ML19344B169 | List: |
| References | |
| WVY-80-117, NUDOCS 8008250691 | |
| Download: ML19344B168 (12) | |
Text
,
Propossd Change No. 89 VERMONT YAN KME NUCLEAR POWER CORPORATION SEVENTY SEVEN GROVE STREET RUTLAND, YERMONT 05701 B.3.2.1 aErty To WVY 80-117 ENGINEERING OFFICE TURNPIKE ROAD WESTBORO, M ASS ACHUSETTS 01581 TELEPHONE 697-366-9019 August 19, 1980 United States Nuclear Regulatory Commission i Washington, D.C.
20555 Attention: Office of Nuclear Reactor Regulation
Reference:
(1) License No. DPR-28 (Docket No. 50-271)
Subject:
Reload 7 Licensing Submittal
Dear Sir:
Pursuant to Section 50.59 of the Commission's Rules and Regulations, Vertnont Yankee Nuclear Power Corporation hereby proposes the following changes to Appendix A of the operating License.
Proposed Change:
The changes are described on Attachment 1.
Revised pages are attached.
Reason for Change:
The reasons for the changes are described on Attachment 1.
The only change directly related to the core reload is the change to the MCPR operating limits. The other changes are updates of the technical specifications per previous commitments or discussions.
Basis for Change:
The basis for the changes are provided in the references listed on.
Safety - Considera tions :
None of the changes are deemed to involve significant unreviewed safety questions.
The change to the definition of " operable" is administrative in nature and is being done at the NRC's request.
8 008250 N
r U.S. Nucle r Ragulatory Commission Pags 2 Attention Offica of Nuclear R;1ctor Regulation August 19, 1980 The change to the Reactor Protection System maximum allowable response time delay from 100 milliseconds to 50 milliseconds is being done to assure operation of the plant within the bounds of the licensing analysis. The discrepancy between the current value and the licensing analysis basis was discovered through an in-depth review of the licensing analysis basis by Vermont Yankee. Administrative controls were immediately placed on the plant to assure that the plant would not operate outside the bounds of the licensing analysis when the discrepency was first discovered. The NRC was notified, and, as a consequence, IE Circular No. 80-08 was issued to all operating plants. This circular contains a caution that any reduction in the response time delay in technical specifications should include an evaluation of the capability to accurately measure that delay time. Scram response time delays are measured on a strip chart recorder, which is operated at a speed of 20 cm/sec and which has an accuracy of + 0 5%. Typically, the delay time is 30 ms, which corresponds to 6 mm on the chart. The maximum allowable delay of 50 ms would correspond to 10 mm on the chart. These methods are judged to give sufficient accuracy for measuring the scram time delay.
The wording change to Specification 3 3B.4 reflects the currently approved licensing basis for evaluating the control rod drop accident, thus it does not involve an unreviewed safety question.
The changes to the MCPR operating limits are based
'.n the reload licensing analysis results, as described in the enclosed report. The changes reflect a continuing trend towards higher MCPR operating limits as core average exposure increases and as more retrofit fuel is introduced. The most significant change in the licensing analysis was an increase in peak transient pressures caused by the introduction of the recirculation pump trip feature.
However, the overpressurization analysis. Section 12 of the enclosed report, shows that the ASME Code requirements of 110% design pressure are still met.
Also, Appendix C of the enclosed report shows that the margin-to-spring safety valves requirements are also met. Additional evaluations are in progress to determine if changes are necessary to Specification 3 6D.1, which currently requires a reactor power reduction to 95% when one relief valve is inoperable. Supplemental information will be provided if additional changes are required.
This submittal has been reviewed by the Vermont Yankee Nuclear Safety Audit and Review Committee.
Fee Determination:
This proposed change requires an approval that involves a single safety issue and is deemed not to involve a significant hazards consideration. For these resons, Vermont Yankee Nuclear Power Corporation proposes this change as a Class III Amendment. A payment of $4,000.00 is enclosed.
Schedule of Change:
This proposed change should be approved no later than November 14, 1980, which is the currently planned startup date for Cycle 8.
U.S. Nuclear Ragulatory Commission Page 3 Attention Office of Nuclear Reactor Regulation August 19, 1980 We trust you will find this submittal acceptable; however, should you have any questions, please contact us.
Very truly yours, VERMONT YANKEE NUCLEAR POWER CORPORATION N YY L. H. Heider Vice President COMMONWEALTH OF MASSACHUSETTS)
)ss COUNTY OF WORCESTER
)
Then personally appeared before me, L. H. Heider, who, being duly sworn, did state that he is a Vice P esident of Vermont Yankee Nuclear Power Corporation, that he is duly authorized to execute and file the foregoing request in the name and on the behalf of Vermont Yankee Nuclear Power Corporation, and that the statements therein are true to the best of his knowledge and belief.
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Description of Proposed Changes to Tach. Specs.
Section Location Change Reason 1.0 P.2, Item K Replace the current definition of operable In response to NRC letter, with the following:
D. G. Eisenhut to All Power "A system, subsystem,. train, component Reactor Licensees, dated or device shall be OPERABLE or have April 10, 1980 per commitment OPERABILITY when it is capable of perform-in VY letter, L. H. Heider to ing its specific function (s).
Implicit USNRC, WVY 80-80 dated in this definition shall be the assumption May 22, 1980.
that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubri-cation or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s)."
3.lA P.18 Change "100 milliseconds" to The reload licensing analysis "50 milliseconds" is based on 50 ms RPS delay time, thus this change brings the tech. spec. into conformance with the licensing i
basis, per IE Circular 80-80/
3 3B.4 PP.70 R,71 Replace 3 3B.4 in its entirety with the The maximum rod worth criterion following:
is no longer directly applicable,
" Control rod patterns and the sequence per the approved Generic Reload of withdrawal or insertion shall be Fuel Application, NEDE-240ll-P-A.
established such.that the rod drop accident limit of 280 cal /g is not exceeded".
t Replace the bases for this epecification with I
a reference to the GE Generic Reload Fuel Application.
3 11 Table 3 11-2 Replace the current MCPR operating limits The revised values reflect the P. 180-01 with the enclosed.
Cycle 8 core conditions and the evaluation of limiting transients per the Supplemental Reload Licensing Submittal, Y1003J01 A02, July 1980 (enclosed).
F VYNPS
- G.
Instrument Functional Test - An instrument L.
Operating - Operating means that a system or component functional test means the injection of a is performing its intended functions in its required ~
simulated signal into the instrument primary manner.
sensor, to verify the proper instrument channel response, alarm, and/or initiating action.
M.
Operating Cycle - Interval between the end of one refueling outage and the end of the next subsequent 11. Log System Functional Test - A logic system refueling outage.
functional test means a test of all relays and contacts of a logic circuit from sensor to N.
Peaking Factor - The ratio of the fuel rod heat flux activiated device to insure all components are to the heat flux ef an average rod in an identical operable per design intent.- Where possible, geometry bundle operating at.the average core power.
action will go to completion, i.e.,
pumps will be started and valves opened.
O.
Primary Containment Integrity - Primary containment I.
Minimum Critical Power Ratio - The Minimum integrity means that the drywell and pressure Critical Power Ratio is defined as the ratio of supression chamber are intact and all of the following I
that power in a fuel assembly which is calculated conditions are satisfied:
to cause some point in that assembly to experience boiling transition as calculated by 1.
All manual containment isolation valves on lines application of the GEXL correlation to the actual connecting to the reactor coolant system or assembly operating power.
containment which are not required to be open (Reference NEDO-10958) during accident conditions are closed.
J.
Mode - The reactor mode is that which is 2.
At least one door in each airlock is closed and established by the mode-selector-switch, sealed.
K.
Operable - A system, subsystem, train, component 3.
All automatic containment isolation valves are or device shall be OPERABLE or have OPERABILITY operable or deactivated in the isolated position.
when it is capable of performing its specified function (s).
Implicit in this definition shall 4.
All blind flanges and manways are closed, be the assumption that all necessary attendant instrumentation, controls, normal and emergency P.
Protective Instrumentation Definitions electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are 1.
Instrument Channel - An instrument channel means required for the system, subsystem, train, an arrangement of a sensor and auxiliary c omponent or device to perform Jts function (s) equi pment required to generate and transmit to a are also capable of performing :neir related trip system a single trip signal related to the support function (s).
plant parameter monitored by that instrument channel.
2
VYNPS 2.
Trip Systen - A trip cy= tea means an' arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A trip system may require one or more instrument channel trip signals l
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VYNPS 1
J 3.1 LIMITING CONDITIONS FOR OPERATION 4.1 SURVEILLANCE REQUIREMENTS 3.1' REACTOR PROTECTION SYSTEM 4.1 REACTOR ' PROTECTION ' SYSTEM
. Applicability:
Applicability:
Ii Applies to the operability of plant instrumentation Applies to'the surveillance of the plant.
an<! control systems required for reactor safety.
instrumentation and control systems required for reactor safety.
Objective:
Objective:
To specifiy the limits imposed on plant operation by To specify the type and frequency of surveillance those instrument and control systems required for to be applied to those instrument and control reactor safety.
systems required for reactor safety, f
Speci f ica tion:
Specification:
A.
Plant operation at any power level shall be A.
Instrumentation systems shall be permitted in accordance with Table 3.1.1.
The functionally tested and calibrated as i
system response time from the opening of the indicated in Tables 4.1.1 and 4.1.2,
]
sensor contact up to and including the opening of respectively.
l the scram solenoid relay shall.ot exceed 50
}
milliseconds.
i B.
During operation with a maximum' total peaking B.
Once a day during reactor power operation a
i factor (MTPF) greater than the deeign.value ( A) the peak heat flux and total peaking factor i
cither:
shall be determined and the APRM system gains shall be adjusted by'the ratios given a.
The APRM System gains shall be adjusted by in Technical Specifications 2.1. A.I.a and l
the ratios given in Technical Specifications 2.1.B.
2.1.A.1 and 2.1.B or F
b.
The power distribution shall be changed to reduce the maximum total peaking factor (MTPF) to'or.less than the-design value (A).
18 I
VYNPS 3.3 LIMITING-CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS 2.
The control rod drive housing support system 2.
The control rod drive housing support system
. shall be in place when the reactor coolant system shall be inspected after reassembly 'and the
- is pressurized above a tmospheric pressure with results of the inspection recorded.
fuel in the reactor vessel unless all operable control rods are fully inserted.
3.
While the reactor is below 20% power, the Rod 3.
Prior to control rod withdrawal for startup the -
Worth Minimizer (RWM) shall be operating while Rod Worth Minimizer (RWM) shall'be verified as
..oving controls _ rods except that:
operable by performing the following:
(a)
If af ter withdrawal of at least twelve (a) The Reactor Engineer shall verify that the control rods during a startup, the RWM control rod withdrawal sequence for the Rod fails, the startup may continue provided a Worth Minimizer computer is correct.
second licensed operator verifies that the operator at the reactor console is following (b) The Rod Worth Minimizet diagnostic test the control rod program; or shall be performed.
(b)
If all rods, except those that cannot be (c) Out-of-sequence control rods in each moved with control rod drive pressure, are distinct RWM group shall be selected and the fully inserted, no more than two rods may be annunciatior of the selection errors moved.
verified.
(d) An out-of-sequence control rod shall be withdrawn no more than three notches and the rod block function verified.
4.
Control rod patterns and the sequence of 4.
The control rod pattern and sequence of withdrawal or insertion shall be established such withdrawal or insertion shall be verified to that the rod drop accident limit of 280 cal /g is comply with Specification 3.3.B.4.
not exceeded.
70R J
VYNPS 3.3 LIMITING CONDITIONS FOR OPERATION 4.3 SURVEILLANCE REQUIREMENTS f
i 5.
Control rods shall not be withdrawn for startup 5.
Prior to control rod withdrawal for startup or or refueling unless at least two source range during refueling, verification shall be made that channels have an observed count rate greater than at least two source range channels have an or equal to three counts. per second.
observed count rate-of at least three counts per second.
6.
During operation with limiting control rod 6.
When a limiting control rod pattern exists, an patterns either:
instrument functional test of the RBM shall be performed prior to withdrawal of the designated (a) Both RBM channels shall be operable; or
- od(s) and. daily thereaf ter.
(b) Control rod withdrawal shall be blocked; or (c) The operating power level shall be limited so that the MCPR will-remain above the fuel cladding integrity safety limit assuming a single error that results. in complete withdrawal of any single operable control rod.
71
~_
I VYNPS 3.3 (cont'd)
[
B.
1.
Control rod dropout accidents as discussed in the FSAR can lead to significant core damage.
If coupling integrity is maintai'ed, the possiblity of a rod dropout accident is eliminated.' The overtravel position feature l
providea a positive check as only uncoupled drives may reach this position. Neutron instrument 0 tion response to j
rod movement provides a verification that the rod is following its drive.
2.
The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the
(
extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal single withdrawal increment, will not contribute to any damage of j
the primary coolant system. The design basis is given in Subsection 3.5.2 of the FSAR, and the design evaluation i
is given in Subsection 3.5.4 This suppo.c is not required if the reactor coolant system is at atmospheric
]
pressure since there would then be no driving force to rapidly eject a drive housing.
3.
In the course of performing normal startup and shutdown procedures, a pre-specified sequence for the withdrawal or insertion of control rods is. followed. Control rod dropout accidents which might lead to significant core damage, cannot occur if this sequence of rod withdrawals or insertions is followed. The Rod Worth Minimizer restricts withdrawals and insertions to those listed in the pre-specified sequence and provides an additional 3'
check that the reactor operator is following prescribed sequence. Although beginning a reactor startup without having the RWM operable would entail unnecessary risk, continuing to withdraw rods if the RWM fails-subsequently is acceptable if a second_ licensed operator verifies the withdrawal sequence. Continuing the startup increases core power, reduces the rod worth and reduces the consequences of dropping any rod.
Withdrawal of rods for j
testing is permitted with the RWM inoperable, if the reactor is subcritical and all other rods are fully j
inserted. Above 20% power, the RWM is not needed since even with a single error an operator cannot withdraw a i
rod with sufficient worth, which if dropped, would result in anything but minor consequences.
4.
Refer to section 5.5.1 of NEDE 24011P-A, latest revision, " Control Rod Drop Accident Evaluation".
4 i
1 1
1 l
76 i
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UYNPS 5.
The Source Range Monitor (SRM) system has no acram functions.
It does provide the operator with a visual Indiention of neutron level. The consequences of renetivity accidenta are n ' function of the initial neutron flux. The requirement of at lennt three counts per second annurca that any transient, should it occur, begins at or above the initial value of 10-8 of rated power used in the nnalyses of transients from cold conditions..One operable SRM channel is adequate to monitor the approach to criticality therefore, two operable SRM's are speci fied for added conservati nm.
6.
The Rod Illock Monitor (RBM) in designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation.
During reactor operation with certain limiting control rod patterns, the withdrawal of a denignated aingle control rod could result in one or more fuel rods with MCPR lean than the fuel cladding integrity safety limit.
During une of such patterns, it is judged that tenting of the RIIM nystem prior to withdrawal of such rods will provide added nasurance that improper withdrawal does not occur.
It in the renponshility of the Nuclear Engineer to identify these limiting patterns and the designated roda cither when the patterns are initially established or as they develop due to the occurrence of inoperable control rods.
77 J
r
=7 VYNPS Table-3.11-2 MCPR OPERATING LIMITS Value of "N" Fuel Type in Exposure Range RBM Equation (1) 8x8 8x8 P8 x 8R BC+ to EOC-2 GWd/t 42%
1.21 1.26 1.27 41%
1.21 1.22 1.23 40%
1.21 1.21 1.22 39%
1.21 1.21 1.21 EOC-2 GWd/t to EOC-1 GWD/t 42%
1.26 1.26 1.28 41%
1.26 1.26 1.28
<40%
1.26 1.26 1.28 EOC-1 GWd/t to EOC 42%
1.29 1.29 1.31 41%
1.29 1.29 1.31
<40%
1.29 1.29 1.31 (1) The Rod Block Monitor trip setpoints are determined by the equatior. shown in Tr* 1e 3.2.5 of the Technical Specifications.
(2) The current analysis for MCPR Operating Limits do not include 7 x 7 fuel.
On this basis further evaluation of MCPR operating limits is required before 7 x 7 fuel can be used in Reactor Power Operation.
180-01