ML20203A883
| ML20203A883 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 07/15/1986 |
| From: | Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | Noonan V Office of Nuclear Reactor Regulation |
| References | |
| SBN-1158, NUDOCS 8607170352 | |
| Download: ML20203A883 (7) | |
Text
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SEABROOK STATION
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Engin n ring Offico e't-L 4
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July 15, 1986 Pubbc Service of New HompeNro SBN-1158 T.P.
B7.1.3 NEW HAMPSHIRE YANKEE DIVISION dnitedStatesNuclearRegulatoryCommission Washington, DC 20555 Attention:
Mr. Vincent S. Noonan, Project Director PWR Project Directorate No. 5 References (a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444
Subject:
Control Rod Drop Testing
Dear Sir:
During recent discussions with the Staf f a change to FSAR Section 14.2 was proposed regarding limiting control rod testing to hot full-flow conditions. This would eliminate unnecessary cycling of the reactor coolant pumps for testing at no-flow conditions. As discussed with the Staff this would be consistent with c.urrent reg-ulatory practice and policy.
Accordingly, enclosed please find a revision to FSAR Section 14.2 which is in accordance with the above.
This revision will be incorporated into the FSAR via a future amend-ment.
Very truly yours,
)
W John DeVincentis Director of Engineering Enclosure cc Atomic Safety and Licensing Board Service List 0607170352 060715 PDR ADOCK 05000443 1 I h
ppy Seabrook StationConstructionFieldOffico. P.O. Box 700
- Seabrook,NH00874
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- a Curren, E:quiro Pater J. Mathsws, Mayor H2rson & Weies City Hsil 2001 S. Street, N.W.
Neuburyport, MA 01950 Suite 430 Washington, D.C.
20009 Judith H. Minner Silversate, Gertner, Baker, Sherwin E. Turk, Esq.
, Fine, Good & Misner Office of the Executive Legal Director 88 Broad Street U.S. Nuclear Reguistory Commission Boston, MA 02110 Tenth Floor Washington, DC 20555 Calvin A. Canney City Manager Robert A. Backus, Esquire City Hall 116 Lowell Street 126 Daniel Street P.O. Box 516 Portsmouth, NH 03801 Manchester, NH 03105 Stephen E. Merrill, Esquire Philip Ahrens, Esquire Attorney General Assistant Attorney General George Dana Bisbee, Esquire Department of The Attornef General Assistant Attorney General Statehouse Station #6 Office of the Attorney General Augusta, ME 04333 25 Capitol Street Concord, NH 03301-6397 Mrs. Sandra Gavutis Chairman, Board of Selectmen Mr. J. P. Nadeau RFD 1 - Box 1154 Selectmen's Office Kennsington, NH 03827 10 Central Road Rye, NH 03870 Carol S. Sneider, Esquire Assistant Attorney General Mr. Angie Machiros Department of the Attorney General Chairman of the Board of Selectmen One Ashburton Place, 19th Floor Town of Newbury Boston, MA 02108 Newbury, MA 01950 Senator Gordon J. Humphrey Mr. William S. Lord U.S. Senate Board of Selectmen Washington, DC 20510 Town Hall - Friend Street (ATTN Tom Burack)
Amesbury, MA 01913 Richard A. Hespe, Esq.
Senator Gordon J. Humphrey Hampe and McNicholas 1 Pillsbury Street 35 Pleasant Street Concord, NH 03301 Concord, NH 03301 (ATTN Herb Boynton)
Thomas F. Powers, III H. Joseph Flynn, Esquire Town Manager Office of General Counsel Town of Exeter Federal Eeergency Management Agency 10 Front Street 500 C Street, SW Exeter, NH 03333 Washington, DC 20472 Brentwood Board of Selectmen Paul McEachern, Esquire RFD Dalton Road Matthew T. Brock, Esquire Brentwood, NH 03833 Shaines & McEachern 25 Maplewood Avenue Gary W. Holmes Esq.
P.O. Box 360 Holmes & Ells Portsmouth, NH 03801 47 Winnacunnet Road Hampton, NH 03842 Robert Carrigg Town Office Mr. Ed Thomas Atlantic Avenue FEMA Region 1 North Hampton, NH 03862 442 John W. McCormack PO & Courthouse Boston, MA 02109
SBN-1158 l
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SB 1 & 2 Amendment 58 FSAR April 1986 r,
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a.
The low -power psuedo-rod-ejection test will be deleted for Unit 2.
(Appendix A, Section 4.c) b.
The power coef ficient measurement for Unit 2 will consist of l
a single measurement at approximately 75% power.
(Appendix A, l
Section 5.a.)
- 12. Vibration levels of the reactor coolant system and piping reaction 1
to transient conditions are measured during hot functional testing l
(Appendix A.2.f.)
f 46 j
- 13. Evaluation of rod scram times for scrama that occur during power j
ascension will not be performed since no practical method for obtaining this data exists for a Westinghouse PWR.
(Appendix A, 1
)
Section 5.h).
l 5(e
- 14. The static rod drop test will not be performed at Seabrook.
i Performance of this test at other facilities has resulted in j
abnormally high power tilts and large Xenon oscillations and may increase the risk of fuel failure.
Performance of this test at i
plants similar to Seabrook has provided ample data to demonstrate j
that Westinghouse computer codes are able to adequately predict core thermal and nuclear parameters for RCCA misalignments up to and including full insertion of a single high worth rod.
In t
i addition, following performance of this test a Catawba, INPO has recommended that utilities delete this test from their startup 3
j programs.
(Appendix A, Section 5.f).
i
- 15. The psuedo-rod ejection test will not be performed at greater than 10% power at Seabrook.
Performance of this test may c mit in violation of the Technical Specification limits on pe ung factor.
Since the accident analysis for Seabrook shows the atpower ejected j
rod worth and power peaking factor are bounded by the zero power case, the calculational model will be verified during the pseudo-1 rod-ejection test at zero power.
( Appendix A, Section 5.e).
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Regulatory Guide 1.68.2. Rev. 1 i
Inicial Startup Test Program to Demonstrate Remote Shutdown Capability for J
Water-Cooled Nuclear Power Plants I
l Since the remote shutdown mode of operation is designed to handle a substantial decay hear. Load, operation of the residual heat removal system i
from the remote shutdown panels during the initial test program does not offer suf ficient reassurance that Technical Specifications cooldown ilmits j
would not be vloisted while performing the cold shutdown demonstration as j
described in Regulatory Guide 1.68.2.
i i
j The remote shutdown capability of Seabrook Station will be demonstrated in j
cccordance with the intent of Regulatory Guide 1.68.2 except as follows:
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14.2-7a
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ENCLOSURE TO SBN-1158 (Continued) i j
INSERT "A"
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onto page 14.2-7a I
4 16.
Control rod scram times will be measured at hot full-flow l
conditions only. Testing at no-flow conditions would re-quire unnecessary and undesirable cycling of the reactor coolant pumps in order to be in compliance with the Technical Specifications.
In addition, the hot full-flow condition is the limiting condition required by Technical Speci fications. ( Appendix A, Section 2.b).
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4 SBN-1158 i
SB 1 & 2 Amendment 44 FSAR February 1982 i
the possibility of mechanical damage, define the criteria for stopping fuel loading, containment evacuation, and emergency boration, define the conditions which must exist for fuel loading to proceed, define responsibility and cuthority of personnel involved in the operation, and provide for fuel and i
core component accountability and status.
14.2.10.2 Initial Criticality 44 Upon completion of fuel loading, the reactor upper internals and the pressure vessel head are installed. Mechanical and electrical tests are performed on the control rod drive mechanisms. These tests include a complete opera-l tional checkout of the mechanisms and calibration of the individual rod position indication system.
Tests are performed on the reactor trip circuits to verify manual trip opera-tion of control assemblies. The control assembly drop times are measured for each control rod assembly at beeb hot _..?
df " _2 conditions 4 t
_a
--..-.__.a Qt gy During control rod drive mechanism testing, the boron concentration in the coolant-moderator is maintained such that the shutdown margin requirements specified in the Technical Specifications are met During individual control i
j assembly or control bank movement, source range instrumentation is monitored for unexpected changes in core reactivity. A functional check is made of the moveable incore detector system and a leak test of the reactor coolant l
system is performed. Just prior to the approach to criticality, a functional test of the nuclear instrumentation is conducted, including verification that the high flux scram setpoint is set at a low value.
I Initial criticality is achieved with the reactor at normal operating tempera-ture and pressure by a combination of shutdown and control bank withdrawal I
and reactor coolant system boron reduction.
Inverse count rate ratio monitor-ing, using data from the normal plant source range instruments, is used as an indication of the proximity and rate of approach to criticality.
Inverse count rate ratio data are plotted as a function of rod bank position I
during rod motion and as a function of primary water addition during reactor coolant system boron concentration reduction.
l Initially, the shutdown and control banks are withdrawn in the normal with-I drawal sequence leaving the last withdrawn control bank inserted far enough l
in the core to provide ef fective control when criticality is achieved.
i The boron concentration of the reactor coolant system is then reduced by 1
the addition of primary water. Criticality is achieved during boron dilution I
or by subsequent rod withdrawal following boron dilution. The rate of primary water addition and, therefore, the rate of approach to criticality may be reduced as the reactor approaches criticality to ensure that ef fective control is maintained. Throughout this period, samples of reactor coolant are obtained and analyzed for baron concentration.
14.2-12
SBN-1158 SB 1 & 2 d
Amendeest 48 FSAR January 1983 TABLE 14.2-5 i
(Sheet 10 of 53) 7.
ROD DROP TIME MEASUREMENT 1
Objective To determine the drop time of each full length control rod under various plant conditions.
Plant Conditions / Prerequisites Prior to initial criticality, during N hot standby condi-tions with full flow ;;f : f l er -- d 4 * ' -- - in the reactor coolant system.
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Test Method During each of the applicable plant conditions, the drop time for each rod control cluster assembly will be determined. Those control rods whose i
drop times fall outside the two-sigma limit determined from the data for all control rods will be retested at least three times to ensure proper performance.
Acceptance Criteria The rod drop times meet the requirements given in Technical Specifications Section 3.1.3.4.
46
SBN-1158 SB 1 & 2 Amendment 56 FSAR November 1985 TABLE 14.2-5 (Sheet 25 of 53) 22.
NATURAL CIRCULATION TEST Objective To verify the ability of the reactor coolant system to remove heat by means of natural circulation.
Plant Conditions / Prerequisites l
The plant is critical at low power.
Test Method i
At hot no-flow conditions (i; :: j-..;;;..... o
..l 1 ;p ;;.;...., T. 1. 1,.2-;
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7' the pressurizer heaters will be turned off and data will be collected l
co determine a depressurization rate.
With the plant at steady state low power conditions (approximately 3%), the reactor coolant pumps will be tripped. This test will determine the length of time necessary to stabilize natural circulation and will demonstrate the reactor coolant flow distribution by obtaining in-core thermocouple maps.
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Auxiliary spray will be used to partially depressurize the primary plant, and the depressurization rate will be determined. At reduced pressure the effect 64 of changes in charging flow and steam flow on subcooling will be verified.
Data will be collected during the test to verify simulator modeling.
Acceptance Criteria Natural circulation is established and maintained as indicated by stable l
temperature indication.
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