ML20207N419
| ML20207N419 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 01/07/1987 |
| From: | Corbin McNeil Public Service Enterprise Group |
| To: | NRC OFFICE OF ADMINISTRATION (ADM) |
| References | |
| NLR-N86199, NUDOCS 8701140202 | |
| Download: ML20207N419 (7) | |
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.g m Pubhc Service Electric and Gas Company Ctrbin A. McNeill, Jr.
Public Service Electric and Gas Company P.O. Box 236. Hancocks Bridge, NJ 08038 609 339-4800 Vice President -
Nuclear January 7, 1987 NLR-N86199 United States Nuclear Regulatory Commission Document Control Desk-Washington, DC 20555 Attention:
Document Control Desk Gentlemen:
INITIAL START-UP TEST PROGRAM CHANGE
' HOPE CREEK GENERATING STATION DOCKET NO. 50-354 In accordance with License Condition 2.C(10) of Facility Operating License NPF-57 dated July 25, 1986 and the provisions of 10 CFR 50.59, Public Service Electric and Gas Company (PSE&G) hereby submits 39 copies of a change made to the Hope Creek
. Generating Station (HCGS) Initial Start-Up Test Program.
The HCGS Initial Start-Up Test Program is described in Chapter 14. 2 of the Final Safety Analysis Report (FSAR).
Attached to this transmittal is a description of, justification for and 10 CFR 50.59 Evaluation supporting the change.
Per the requirements of 10 CFR 50.59, Paragraph (a)(2), the attached Safety Evaluation provides the basis for the conclusion that the attached FSAR revisions do not involve an unreviewed safety question.
f Should you have any questions regarding this transmittal, please j
do not hesitate to contact us.
Sincerely, t
prb Attachments (2) r 11 k
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1/7/07
..Dr.
Thomas E. Murley 2-4 C-Mr.
D.
H.-Wagner USNRC Licensing Project Manager 4
Mr.
R.
W.
Borchardt USNRC Senior Resident Inspecto r 4
Mr. J. M. Taylor Director - Inspection and Enforcement Dr.
T.
E.
Murley Regional Adminstrator, USNRC Region I i-f 1
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ATTACHMENT 1 I.
DESCRIPTION OF CHANGE A change has been made to FSAR Figure 14.2-4 which extends the bottom of the Test Condition 4 window (TC-4) down to 38.5 percent power and appropriately revises the footnote associated with the window.
II.
JUSTIFICATION FOR THE CHANGE TC-4 testing is conducted during the Initial Startup Test Program and consists of testing the plant systems during natural circulation conditions (i.e. recirculation pumps tripped off).
The area of the power / flow operating map where such testing is performed is generally identified in FSAR Figure 14.2-4 and was specifically identified by PSE&G as being "on the natural circulation core flow line - within +0,
-5% of the intersection with the 100% power rod line."
Prior to actually conducting TC-4 testing, test engineers anticipated power would drop to appros imately 50% rated power upon tripping the recirculation pumps.
Hence, in order to maintain an adequate scram avoidance margin as required by procedure (7.5%), not trip the plant by exceeding the APRM Thermal Power Monitor Scram Setpoint (51% of rated power at natural circulation) and still maintain reactor power as high as possible without unnecessarily challenging the safety systems, the control rods were inserted to between the 80% and 95% power rod lines.
When the recirculation pumps were tripped, powe r dropped to 43% of rated power as anticipated and natural circulation testing commenced.
Testing in TC-4 was completed approximately 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> later at which time the power had dropped to 38.5% rated power due to zenon buildup.
As a result of the testing performed in TC-4 and in order to accept the results, a detailed evaluation was required to justify the data obtained and properly revise FSAR Figure 14.2-4 allowing the power to drop to 38.5% rated power with control rods inserted between the 80% and 95% power rod lines (i.e.
"+0,
-20% of the intersection with the 100% power rod line").
This detailed evaluation is summarized in the following paragraphs.
One of the objectives of the testing in TC-4 was to provide the operators with training during natural circulation operation and to adequately characterize the system performance during these conditions.
Some of the testing performed at natural circulation conditions (e.g. recirculation system performance, monitoring of selected process temperatures and monitoring of core performance) is not significantly sensitive to the reactor power level
during the testing.
Therefore, the small reduction in power level reflected in this test does not impact system performance.
In fact, this same argument was utilized by the Clinton Power Station where the TC-4 window was expanded to include the 80% power rod line.
A second objective of TC-4 testing is to demonstrate control system testing.
However, since testing in TC-4 is bounded by testing in TC-3 and TC-5, it is not necessary to repeat the control system demonstrations performed at these other test conditions, and in fact, control system testing in TC-4 has been deleted for this reason at the Clinton Power Station.
Regardless, PSE&G did perform control system testing in order to obtain additional information and data points.
This position is justified since the responses of the pressure regulator and feedwater control systems are primarily sensitive to vessel steam flow which is proportional to power level.
Therefore, testing of the control systems at reduced power levels, although unnecessary to prove system performance, do provide adequate data and can be utilized as a redundant test of the systems.
Another objective of TC-4 testing is to meet the intent of Regulatory Guide 1.68, Appendix A, Paragraph 5 which states that
" appropriate consideration should be given to testing at the extremes of possible operating modes for the facility systems."
Although natural circulation is not an intended operating mode, the plant could be in such a situation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> before plant shutdown is required (Technical Specification 3/4.4.1).
Therefore, stability of the plant was to be demonstrated by performance of various tests, including the pressure regulator tests, in the extreme portion of the High Flux - Low Flow region of TC-4.
In Supplement 6 to the NRC Safety Evaluation Report (NUREG-1048) Appendix S, Modification #7, the NRC staff agreed with PSE&G "that the generic natural circulation and stability have been sufficiently confirmed in recent domestic and foreign commercialized BWRS...however... plant specific confirmatory testing" is still necessary.
Therefore, the testing performed by PSE&G in TC-4 is plant specific in nature and thus the plant specific revisions required to accomplish the testing are justifiable based upon the plant specific information which was obtained.
Finally, the pressure regulator testing (specifically the pressure setpoint changes) at TC-4 provides information on the stability of the system.
Additional information on the stability of the reactor at TC-4 is available by monitoring the neutron flux (both local and average) as required by Technical Specification 3/4.4.1.1.
Pressure regulator testing performed at TC-5, which bounds the least stable portion of the nonnal operating region, provided additional information on the stability of the reactor.
Clinton Power Station in fact has provided sufficient justification to delete pressure regulator testing in TC-4 based upon just such an argument.
Therefore, the testing which PSE&G conducted, although unnecessary, does provide valuable data even at a reduced power level during natural circulation and is therefore acceptable.
In conclusion, the TC-4 test conditions are not specified in Regulatory Guide 1.68.
By performing the testing in TC-4 at 43%
power meets the intent of the guideline without unnecessarily challenging the safety systems and provides additional information, even though such inforaation is readily available from other tests.
All testing scheduled for performance in TC-4 was completed satisfactorily.
III.
10 CFR 50.59 SAFETY EVALUATION The responses to the following questions, and the detailed justification provided above, clearly indicate that the changes shown in Attachment 2 to FSAR Figure 14.2-4 do not represent an unreviewed safety question.
A.
Does the proposed action increase the probability of occurrence or the consequences of an accident or malfunction of equipment related to safety, as previously evaluated in the FSAR?
No, because no changes to the physical plant were made and therefore, no increase in the probability of occurrence or consequences of an accident as previously evaluated occurs.
Tests parformed in the TC-4 window are demonstration tests and are not affected by the change in the power level.
B.
Does the proposed action create the possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR?
No, because no changes in the physical plant were made and no operation outside of previously analyzed regions was performed.
The fact that testing is done at a lower power level does not invalidate the test itself.
C.
Does the proposed action reduce the margin of safety as defined in the basis for any Technical Specification?
No, because the margin of safety is not reduced as no change to the actual Technical Specifications is required and no change to the physical plant occurred.
ATTACHMENT 2
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10 20 30 40 50 60 70 80 90 10 0 11 0 PERCENT CORE F' LOW TEST CONDITION (TC) REGION DEFINITIONS TEST CONDITION NO.
POWER FLOW MAP REGION AND NOTES
,1 BEFORE OR AFTER M AIN GENERATOR SYNCHRONIZATION BETWEEN 5.% AND 20.% THERMAL POWER-WITHIN 310.% OF M G SET MINIMUM OPERATING SPEED LINE IN LOCAL MANUAL MODE.
2 AFTER MAIN GENER ATOR SYNCHRONIZATION BETWEEN THE 45.% AND 75.% POWER ROD LINES BE1 WEEN M-G SET MINIMUM SPEEDS FOR LOCAL M ANUAL AND MASTER MANUAL MODES THE LOWER POWER CORNER MUST BE LESS THAN BYPASS VALVE CAPACITY.
3 BETWEEN THE 45.% and 75 '. POWER ROD LINES -
CORE FLOW BETWEEN 80S AND 100.% OF ITS R ATED VALUE.
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f*NM ON THE NATURAL CIRCU: % TION CORE FLOW LINE -
WITHIN +030F THE IN. ;RSECTION WITH THE 100.% POWER ROD LINE.
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WITHIN +0, 5% OF THE 100 % POWER ROD LINE -
WITHIN 5.% OF THE ANAL) TICAL OF THE LOWER i
LIMIT OF MASTER FLOW CONTROL.
6 WITHIN +0,4% 0 F RATED 100.% POWER -WITHIN
+0,-5% OF RATED 100.% CORE FLOW RATE.
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HOPE CREEK GENERATING STATION Tb MRC SuBmstTN FINAL SAFETY ANALYSIS REPORT WTG N UARY ],I W OPER ATIONAL POWER / FLOW MAP FIGURE 14.2 4 Amendment 1,8/83
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