ML19327C067
| ML19327C067 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 11/06/1989 |
| From: | Feigenbaum T PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NYN-89140, NUDOCS 8911150199 | |
| Download: ML19327C067 (16) | |
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MNew Hampshire Ted C. Feigenbaum
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Chief Operating Officer
. Senior Vice President and j
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NYN-89140 November 6, 1989
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United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Document Control Desk
References:
-Facility Operating License NPF-67 Docket No. 50-443
Subject:
NHY Power Ascension Test Pregram, FSAR Chapter 14 Revisions Gentlemen:
New Hampshire Yankee's (NHY) current plans for-power ascension testing necessitate various revisions to Chapter 14 of the Final Safety Analysis-i
' Report (FSAR). : These revisions to FSAR Chapter 14, in the form of annotated i
pages, are enclosed (Enclosure A).
The substantive (non-editorial) FSAR Chapter.14 revisions are summarized as follows:
ST-47 Main Steam Line Isolation Valve Closure Test This test will not be performed.
The objectives of ST-47 are met in l
another test and in typelqualification testing of the Main Steam Isolation Valves (MSIVs) as discussed belcv. The non-performance of
'ST-47 eliminates a deliberate transient on the plant. As indicated previously in FSAR Amendment 48, the dynamic response of the plant to a MSIV closure is bounded by the response of the plant to a turbine trip.
l In ST-38, the unit will be tripped from 100 porcent power, and the plant response will be measured.
s Type / qualification teste performed by Rockwell International on valves 1
similar to the Seabrook MSIVs demonstrate that the Seabrook MSIVs are 4
4 capable of closing under design flow conditions within the required time.
l Ability of Neutton Flux Instrumentation to Detect Control Rod Misalignments q
This test will not be performed.
The Digital Rod Position Indication (DRPI) System provides accurate and sensitive indication of control rod C
position.
The DRPI has been previously tested. Incore flux instru-mentation may be used as an alternative means to detect control rod position.
The Seabrook incore flux instrumentation is identical to instrumentation utilized in all Westinghouse four loop plants since Indian Point 2.
The capability and sensitivity of this instrumentation p
has been demonstrated numerous times.
8911250199 891106 PDR ADOCK 05000443
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PDC New Hampshire Yankee Division of Public Service Company of New Hampshire g
P.O. Box 300
- Seabrook, NH 03874
- Telephone (603) 474 9521 1
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United States Nuclesr Regulatory Commission November 6,.1989 Attention:
Document Control Desk Page 2 ST-22 Natural Circulation Testing This test will be performed in HODE 3 utilizing decay heat to demonstrate natural circulation. The performance of this test using decay heat instead of heat generated by a critical reactor provides more stable plant conditions during testing. Additionally, this test will not include primary system depressurization rate measurements, charging and steam flow variations to determinr. subcooling effects or primary system pressure reductions to verify subcooling monitor performance. The above changes to ST-22 eliminates the potential for esetain transients on the plant.
Enclosure B to this letter provides further discussion and justification for the revisions to the FSAR Chapter 14.
The revisions to the initial test program have been evaluated pursuant to 10CFR50.59. The evaluation determined that the initial test program revisions do not introduce an unreviewed safety question.:
I The enclosed (Enclosure AS FSAR Chapter 14 revisions will be included in the next FSAR amendment.
Should you have any questions.regarding the NHY Power Ascension Test Program. FSAR Chapter 14 Revisions, please contact Mr. Robert E. Sweeney in our.Bethesda Office, at (301) 656-5100.
Very truly yours, b
TedC.Feig[enbaum r
Enclosures l
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.i United States Nuclear Regulatory Commission Novembcr 6, 1989 Attention: Document Control Desk Page 3 cci Mr. William T. Russell Regional Administrator United States Nuclear Regulatory Commission Region I 475 Allendale Road C
King of Prussia, PA' 19406 Mr. Vjctor Nerses, Project Manager Project Directorate I-3 United States Nuclear Regulatory Commission l
Division of Reactor Projects Washington, DC-20555 l
Mr. Antone C. Cerne NRC Senior Resident Inspector p
i P.O. Box 1149' Seabrook, NH. 03874 l
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ENCLOSURE A TO NYN-89140 4
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.May Reculatorv Guide 1.11. Re v. O Preoperat tonal Testing o f Redundsat Electric Power Systems to Verify Proper Load Group Asseenments Seaorack Station coaforms with the recommend.ations of Regulatory Guice 1.LI.
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Regulatora Guide 1.52. Rev. 2 Destgn, Testing and Maintenance Criteria for Engineered Safety Feature Almosphere Cleanup System Air Filtration and Absorption Units of Light Water Cooled Nuciaar Power Plants A detailed discussion on the degree of conformance to Regulatory Guide 1.52 is found in Section 6.5.1.
Regulatorv Guide 1.68. Rev. 2 Initial Test Programs for Water-Cooled Nuclear Power Plants The initial test program for the Seabrook Station will be conducted in accordance with the intant o f Regulatory Guide 1.68 except for the items specified below:
1.
During the preoperational test program. no practical method exists to vary system voltage to ob: sin maximum and minimum design voltages.
The intent of the requirement to demonstrate that the emergency loads can start and operate with the maximum and minimum design voltage available will be met by testing the emergency loads under plant light load conditions to simulate the maximum practically obtainable voltage and under plant heavy 1 ad conditions to simulate h
the minimum practically obtainable voltage. The results of this testing will be compared to the station "oltage study to verify the adequacy of the analytical model.
( Appendix A, Cection 1.g.2).
4r 2.
During the power ascunsior, testing phase, tests will be scheduled such that the safety of the plant will not be dependent on the performance of an untested system or feature.
Power ascension testing will be performed at power plateaus of ap,roximately 30%,
50%, 75% and 100%.
It is required that testing be performed at 30% rather than 25% because individual system stability is increased at 30% (e.g. feedwater system), this allows comparison steady-state conditions with the design at low power. Westinghouse-supplied plants have historically conducted tests at 30% and, therefore, genetic data is available for review and comparison.
3.
Throughout core loading and precritical tests, utdown margin will be verified by periodic sampling of core coolant and verification that boron concentration is maintained at or above the Technical Specification concentration limit for re fueling conditions.
(.*.ppe nd i x A, Seetion 2.a) 4.
Control rod runback and partial scram featares are not used in the Seabrook Station design and, therefore, will not be tested during power escalation.
(Appendix A, Section 5.j.)
5.
A demonsLration of the capability of systems and components to j~
remove residual heat or decay heat from the Reactor Coolant System will be performed during power ascension testing only if not performed 14.2-6
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+4 FSAR April 1986 during hot functional or low power tests.
(Appendix A, Section 5.1.)
- 6.. The failed fuel detection system is not applicable to the Seabrook design and, therefore, will not be tested during power escalacion.
(Appendix A, Section 5.q.)
7.
The integrated control system and the reactor coolant flow control s^
system are not applitable to the Seabrook Station design and there-fore, will not be tested during power escalation.
The Startup and Emergency Feedwater Control Systems and the Steam Pressure Control Systems are only used in the hot shutdown, hot standby or low power operating modes.
These systems can not be tested during power ascension.
(Appendix A, Section 5.s.)
O.
A demonstration of the dynamic response of the plant to a loss of or bypassing 'of a feedwater heater (s) will not be performed.
As shown in Section 15.1.1, the transient resulting from the severe case of feedwater temperature reduction initiated by a most single failure or operator error is similar to, but of a lesser magnitude than the excessive load increase (load swing).
The f
load swing test will be performed at several major plateaus.
4 9.
As shown in Subsection 15.2.3 and 15.2.4, dynamic response of the plant to a MSIV closure is bounded by the response of the plant to the turbine trip event.
Plant response to a turbine trip will be demonstrated durine cerformance of ST-38, unit trin from 100 M po g The ability of the MSIVs to close under steam flow percent will be demonstrated by automatic closure of all main steam line isolation valves at 30 percent power during the performance of Si -4 7.
Apperdix A.
Section 5.m.m.)
s ndt be performed.
10.
The ability of the movable incore neut on flux instrumentation to 1
i detect control rod misalignments will (tte demonstrated at 50'. power) 4 l
M The excore neutron flux instrumentation ts not designed to detect a local condition such as a misaligned RCCA but rather a
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more global anomalous core condition. The movable incere flux
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instrumentation is not intended to specifically detect a misaligned y
control rod, but may be able to confirm an RCCA misalignment initially detected by the rod position indication system. The individual
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rod position indication system is the primary means for detecting I
g RCCA misalignments. Since at 100% power the control rods are y
ersentially withdrawn, individual rod worth is such that the ability h
of the movable incore instrumentation to detect a control rod misalignment is limite_sl./Therefote, data on the movablo incere instrumentatinn characraristics over a range cf control rod insertion will be obtained during a control rod misalignment test at 50% power.J( Appendix A, ~Sec. 5. L ).
11.
Since Units 1 & 2 are essentially identical, the below listed tests, which will be performed on Unit I solely to verify the adequacy of calculational models, will either not be performed or reduced in scope during the Unit 2 inicial startup program.
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The basis for this deletion is that the original requirement was imposed to demonstrate alternative instrumentation capabilities in detecting,a misaligned control rod. With the advent of the Digital Rod Position Indication (DRPI) system, the need for accurate and sensitive alternative indications has been essentially eliminated.
In any case, the distribution and number of the incore and execre flux instrumentation has not been
'L changed and is identical to all Westinghouse four loop plants since Indian Point 2.
Since that time, the capability and sensitivity of of the excore and incore flux instrumentation has been demonstrated numarous times.
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FSAR hpril 1986 1
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(Sheet 2 of 53)
Index.
Title i,<'
Sheet V
22.
Natural Circulation Test 25 a
- 23.. Dynamic Automatic Steam Dump Control 26 H.
24 Automatic Reactor Control 27 f 25. Automatic Steam Generator Level Control 28 2 6.- *hermal Power Measurement and Statepoint Data Collection 29 jk, 27.
Startup Adjustments of Reactor Control System 30 7-
-28.
Calibration ~ of Steam and Feedwater Flow Instrumentation 31
- 29. Core Performance Evatuation 32 i
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Power Coef ficient Measurement 33 31.
(Deleted in Amendment 34 l
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Shutdown from Outside the Control P.com 36 34.
Load Swing Test 37
- 35. (De:h:AJ in ^-- ":::: 57} 4 Ly LN hb b b
38 55
'36., Axial Flux Difference Instrumentation Calibration 39 t
- 37.. Steam Generator Moistura Carryover Measurement 40 38.
Ur.i t Trip from 100 Percent Power 41
- 39r 31ackout Te LW O
- W44 M 42
!$A 40.
NSSS Acceptance Test 43 41.
Radiation Survey 44 1~
42.
Water Chemistry Control 45 j
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Process Computer 46 l-4e l,.
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..f FSAR January 1983 TABLE 14.2-5
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Index Title Sheet
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Vibration and'1.oose Parts Monitor 47 4
- 45.
Process and Efflueet Radiation Monitoring System 48 46.
Ventilation System Operability Test 49
.47.
eam Line Isolation Valve Closure Tes (D M 6 50
- 48. ' Turbine Generator Startup Test a m a s/h o
51 49.
Circulating Water System Thermal-Hydraulic Test 52 e
- 50.c Mova' ble Incore Detector System 53 96
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(Sheet 25 of $3) 22.
NATURAL CIRCULATION TEST j
objeeeive To' verify the ability of the reactor coolant system to remove heat by means i
of natural circulation.
l Plant Conditions /Prereouisites The plant is tical,hlowpow, del J, Q f gg-[Q M
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Test Method t
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At. hot no-flow conditions the prwssurizer hesters will be turned off and) data will be collected to determine a depressurization rate.
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- With the plant at (teady state low oweg hpproximately3" the teactor coolant pumps will be tripped. This test will determine the length of l
time necessary to stabilize natural circulation ar.d will demonstrate the
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teactor coolant flow distribution by obtaini'ng in-core _.hermocouple maps.
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Auxiliary spray will be used to partially depressurize -he primary plant, and the depressurization rate will be detetsined. At reduced pressure the effect u
-of changes in charging flow and steam flow on subraalhe 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
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temperature indication.
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TABLE 14.2-5 (Sheet 42 of 53)
- 39. k ATION BLACKOUT TEST 4.05S M 4MSM #dMR 27
- Objective To demonstrate starting of emergency diesels and proper sequencing of loads following a main generator trip without an available source of offsite power.
Plant Conditions / Prerequisites The plant is at a stable power level of equal to or greater than 10% power.
Test Method Generator output breakers will be tripped resulting in a reactor trip with no offsite power available. The starting of the emergency diesel generators.
and overall plant response trill be monitored.
The loss of offsite power will be maintained long ennugh for plant systems to stabilize (at least 30 minutes or longer).
Acceptance criteria The diesel generators reach rated voltage and frequency within the limits specified in Technical Specifications 4.8.1.1.2.a.
The Emergency Power l
Sequencers function as described in FSAR Section 8.3.
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Amencment a5 FSAR January 1983
.c TABLE 14.2-5 (Sheet 50 of 53) 47.
MAIN STEAM LINE ISOLATION VALVE Cl.0SURE TEST Objective To demonstrate the ability of each main steam isolation valve (MSIV) to close under steam flow.
j Plant. Conditions /Prerecuisites This test will be initiated from steady state conditions at the 30%. power plateau.
Test Method All MSIV's will be closed simultaneously while the plant response is observed.
Selected plant parameters will be recorded including MSIV response
- times..
Acceptance Criteria MSIV closure times are consistent with the requireme.,s contained in Technical Specification 4.7.1.5.
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ENCLOSURE B TO NYN-89140
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U New Hampshire Yankee November 6, 1989 c
e FSAT " action 14.2 has been annotated to reflect revisions to the initial ter' predram and includes a notation in the form of a numbered triangle which correlates to the following numbered basis for each change:
1.
Page 14.2-6, correction of a typographical error.
2.
Page 14.2-7, Item 9, delete the requirement to demonstrate the ability of the MSIV to close under steam flow.
Requirements for stroke testing main steam isolation valves under steam flow conditions were originally established in USNRC Regulatory Guide 1.68. Revision 2, Initial Test Program for Water-Cooled Nuclear Power Plants.
Stroke testing under flow conditions is required at 25Z power and dynamic response testing for the case of closure of all main steam isolation valves is required at 100%
power.
i Acceptable stroke testing of the MSIV's has been demonstrated during hot functional testing in pre-operational test PT-13.
Regulatory l
Guide 1.68 allowed this stroke testing to be performed at less than L
25% power.
I Reflecting the response to RAI-640.49, FSAR Amendment 48 included an I
exception to Regulatory Guide 1.68 stating, ' dynamic response of the plant to a MSIV closure is bounded by the response of the plant to the turbine trip event".
This is tested under ST-38, Unit Trip from 100% Power. It is noted that the MS1V's do not close during ST-38.
l It is the turbine stop valves which close during ST-38, yielding a l
plant dynamic response similar to the closure of the MSIV's at 100%
l power.
l Although not required by Regulatory Guide 1.68, type / qualification l
tests were performed by Rockwell International on similar valves, l
i and demonstrate that the valves are capable of closing under design flow conditions within the required time.
The report which provides the results of these tests is filed at NHY as foreign print FP-23676, " Report RAL 7068 Generic Qualification Report for size 30 x 24 x 30 Figure 1911 (WCC) BGaddPQTY".
RAL 7068 includes a calculation and comparison of maximum stresses between the Seabrook HSIVs and the tested valves, confirming that stresses on the Seabrook valves are less than the stresses on the generic valves.
Isolation capability and response time were demonstrated for higher differential pressure than Seabrook would experience while simultaneously subjected to pipe and earthquake loadings. After demonstrating isolation capability and response times, seat leakage tests were performe:d demonstrating acceptable seating.
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- ~j November 6, 1989 The differences between the Seabrook valve and the tested valves are listed in RAL-7068.
The significant difference is the actuator s
size.
The Seabrook actuator is 182 smaller than the generic valve 1
actuator because the design differential pressure is 232 lower for Seabrook than the generic case. Dimensional factors determining closing forces are virtually identical.
l 3.
Page 14.2-7, Item 10, delete the requirement to demonstrate the ability of the movable incore neutron flux instrumentation to detect control rod misalignment at 50% powar.
The basis for the deletion of this test is that the original requirement was imposed to demonstrate alternative instrumentation capabilities in detecting a misaligned control rod. With the advent of the Digital Rod Position Indication (DRPI) system, the need for' accurate and sensitive alternative indications has been eliminated.
In addition, the distribution and number of the incore and excore flux instrumentation is identical to all Westinghouse four loop plants since Indian Point 2 and the capability and sensitivity of the excore and incore flux instrumentation has been demonstrated numerous times.
Insert A was added to FSAR Section 14.2.7, Item 10 to provide a basisifor exceptions to Regulatory Guide 1.68.
4.
Table 14.2-5, Sheet 2. Item 31, correction for a typographical error.
5.
Table 14.2-5. Sheet 2, Item 37., this test was eliminated in Amendment 58, however, the index was not amended.
6.
Table 14.2-5 Sheet 2. Item 35, this test title was incorrectly eliminated from the index.
7.
Table 14.2-5, Sheet 2, Item 39, change of test title to reflect current test procedure title.
8.
Table 14.2-5. Sheet 3. Item 47, deletion of Main Steam Isolation Valve Closure Test.
See paragraph 4.B.2 above.
9.
Table 14.2-5. Sheet 25, revisions to ST-22, Natural Circulation Test.
The test abstract for ST-22, Natural Circulation Test, has been revised to reflect the following changes in the manner in which this test will be performed.
ST-22 will be performed in MODE 3 utilizing decay heat instead of heat generated by a critical reactor at low power. Recent natural circulation testing at Vogtle utilizing decay heat provided sufficient data to confirm natural circulation flow and heat removal capability.
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November 6, 1989 The primary system depressurization rate with the pressuriser heaters turned off will not be determined.
Instead, the i
pressurizer heat loss rate has been calculated for Seabrook..
Station. ' The calculated value is' comparable to the calculated value for Diablo Canyon as discussed below.
The depressurization rate'of the pressuriser is known to be dependent on ambient heat losses and for Seabrook it was
' determined by Calculation 4.4.13.F-19 Revision 1 that the pressuriser heat loss rate is 2.7'F/hr.
The pressurizer ambient heat loss for a similar plant, Diablo Canyon was calculated'to be approximately (7'F/hr) as described in a final post-test report by the Westinghouse Owners Group. WCAP-11086, dated March 1986.
The depressurization rate with auxiliary pressurizer sprays will not be determined. This is not required by RG1.68 Rev. 2 and is consistent with natural circulation tests performed en plants similar to'Seabrook. The auxiliary pressurizer sprays will be used to control primary system pressure during the revised test.
4 The effects of variations of charging flow and steam flow on subcooling will not be determined.
This is not required by Regulatory Guide 1.68 Rev. 2 and is consistent with natural circulation tests performed on plants similar to Seabrook.
In addition, certain commitments made in response to RAI 640.51 Item i
4.t. are effected as follows:
Fixed incore flux detectors will not be utilized in this test because the reactor will remain ouberitical.
Reduced pressure will not be experienced in this test as discussed i
above, however, the subcooling monitor will be utilized and its performance verified under the natural circulation conditions of this test.
Fifty percent of the available licensed operators will not be L
witnessing the test. Current Operator reque.lification training includes extensive coverage of natural circulation events in accordance with Emergency Operating Procedures. Data obtained during this test will be used to evaluate simulator modeling.
10.
Table 14.2-5, Sheet 42, change of test title to reflect current test procedure title.
11.
Table 14.2-5, Sheet 50, deletion of Main Steam Isolation Valve Closure Test.
See paragraph 4.B.2 above.
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