ML20236N153
| ML20236N153 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 11/04/1987 |
| From: | George Thomas PUBLIC SERVICE CO. OF NEW HAMPSHIRE |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NYN-87127, NUDOCS 8711160042 | |
| Download: ML20236N153 (14) | |
Text
- _ _ _
George S. Thomas j
Vice President-Nuclear Production i
d November 4, 1987 New Hampshire Yankee Division NYN-87127 l
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United States Nuclear Regulatory Commission Washington, DC 20555 i
Attention:
Document Control Desk
References:
(a) Facility Operating License NPF-56, Construction Permit CPPR-136, Docket Nos. 50-443 and 50-444 (b) PSNH Letter (SBN-1090), dated June 5, 1986,
" Steam Line Break Evaluation", J. DeVincentis to the USNRC
Subject:
Main Steam Line Break Outside Primary Containment Gentlemen:
In Reference (b), New Hampshire Yankee (NHY) committed to perf orm and submit to the NRC an analysis to verify that 1.3% Ak/k SHUTDOWN MARGIN was sufficient to meet the f uel design basis f or End-of-Cycle conditions during a postulated main steam line break. Pursuant to this commitment, enclosed please find the steam line break core response anal-ysis for Seabrook Station (Enclosure 1).
In brief, the results of this
. analysis demonstrate that the core response analysis provided in FSAR Section 15.1.5 bounds this new analysis and, as such, the conclusions in the FSAR are unaffected. Therefore, NHY requests that the results of the
'Staf f's review be reflected in the next supplement to the Seabrook Station SER.
In Reference (b), New Hampshire Yankee also committed to an interim modification of the Seabrook Station Technical Specification to require a 3.8% Ak/k SHUTDOWN MARGIN in MODES 1, 2, and 3 until a plant specific reanalysis could be completed and approved.
On the basis that the anal-ysis provided in Enclosure 1 assumes credit for only the standard 1.3%
Ak/k SHUTDOWN MARGIN, we request a change to the Seabrook Station Techni-cal Specifications as shown in Enclosure 2.
This change will remove the interim requirement for 3.8% Ak/k SHUTDOWN MARGIN and return to the stan-dard 1.3% Ak/k.
8711160042 871104 DR ADOCK 05000443 PDR i
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P.O. Box 300. Seabrook, NH 03874. Telephone (603) 474-9574
i Unitqd States Nuclear. Regulatory Commission NYN-87127
. Attention: Document' Control Desk Page 2 i
Should you have any questions concerning our response, please con-tact the NHY Bethesda Licensing Office (Mr. R. E. Sweeney) at (301) 656-6100.
Ve ry t ruly yours,
Ge e S. Thomas Enclosures cc:
Mr. Victor Nerses, Acting Director PWR Project Directorate I-3 i
Division of Projects I/II h'
U. S. Nuclear Regulatory Commission Washington, DC 20555 Mr. William T. Russell.
Regional Administrator U. S. Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406 Mr. A. C. Cerne NRC Senior Resident Inspector Seabrook Station Seabrook, NH 03874 l
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ENCLOSURE 1 TO NYN-87127 STEAM LINE BREAK CORE RESPONSE ANALYSIS ASSUMING 1
CONSEQUENTIAL FAILURES DDE TO SUPERHEATED STEAM i
BACKGROUND:
A Main Steam Line Break located in the Main Steam and Feedwater pipe chase could result in the reopening of the two affected Main Steam Iso-lation Valves (MSIVs) because the presence of superheated steam ex-ceeds environmental qualification conditions.
It is assumed that the presence of superheated steam would result in the consequential reopening of two MSIVs and therefore, would result in the blowdown of two Steam Generators. The blowdown of two Steam Generators is an event which has not been previously evaluated. in Chapter 15 of the Seabrook Station FS AR.
New Hampshire Yankee (NHY) has provided an evaluation (Reference 1) demon-strating that the potential reactivity added to the core by the consequen-tial opening of two MSIVs is negligible when compared with the negative reactivity available by including the actual Cycle 1 SHUTDOWN MARGIN (3.8% Ak/k) instead of the standard (1.3% Ak/k) value.
NHY therefore committed to an interim modification of the Seabrook Station Technical Specifications to require a 3.8% Ak/k SHUTDOWN HARGIN in Modes 1,2 and 3.
This concern was documented by the NRC in Section 3.11.3.3.2 of SER Supple-ment 5 (Reference <4), and found acceptable utilizing a SHUTDOWN MARGIN of 3.8% A k/k.
NHY also committed to perform a plant specific analysis of the core re-sponse to a Main Steam Line Break assuming the consequential reopening of two MSIVs and taking credit for the standard SHUIDOWN MARGIN of 1.3%A k/k. The following evaluation is intended to resolve the remaining concerns relative to this issue.
IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION:
The steam release resulting f rom rupture of a main steam line would cause an initial increase in steam flow which would then decrease during the accident as steam pressure falls. The energy removal f rom the Reactor Coolant System (RCS) causes a reduction of coolant temperature and pressure.
In the presence of a negative moderator temperature coefficient, the cool-down results in an insertion of positive reactivity.
If the most reactive Rod Cluster Control Assembly (RCCA) is assumed stuck in its f ully withdrawn position af ter reactor trip, there is an increased possibility that the core will become critical and return to power.
A return to power following a steam line rupture is a potential problem mainly because of the high power peaking f actors which exist assuming the most reactive RCCA to be stuck in its fully withdrawn position. The core is ultimately shutdown by the boric acid delivered to the Reactor Coolant System by the Safety Injection System.
The analysis of a main steam line rupture is presented in the Seabrook FS AR to demonstrate that the following criteria are satisfied:
Assuming a stuck RCCA with or without of f site power, and assuming a single f ailure in the engineered saf ety features, the core re-. - -. -
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' ENCLOSURE 1 TO NYN-87127 4
(Continued) l STEAM LINE BREAK CORE RESPONSE ANALYSIS ASSUMING '
CONSEQUENTIAL FAILURES DUE TO SUPERHEATED STEAM i
i i
. mains in place ' and intact. Radiation' doses do not exceed the i
limits of 10CFR100.
The FSAR analysis shows that no DNB occurs for _any rupture assuming the most reactive RCCA stuck in its fully withdrawn position. The DNB design basis for.the hypothetical steam line break event has f
been modified.
The pressures for this event fall into the low i
pressure range (500-1000 psia) where the W-3 based DNB correlation l
Lis used for the DNB Ratio (DNBR) design baois with a 1.45 limit l
DNBR. The justification for this new design limit for low pressure applications of the W-3 correlation has been documented by a Westing-house submittal to the USNRC (Reference 2).
l I
, The main steam line rupture is the most limiting cooldown transient and is analyzed at zero power with no decay heat.
Decay heat would retard the cooldown thereby reducing the return to power. 'A detailed analysis of this transient with the most limiting break size, a double-ended rup-ture, is presented in Seabrook TS AR Section 15.1.5.
The following features provide the protection for a steam line rupture:
i 1.
Safety Injection System actuation from any of the following:
a.
Two out of three low steam line pressure signals in any one loop.
b.
Two out of four low pressurizer pressure signals.
c.
Two out of three high-1 containment pressure signals.
2.
The overpower reactor trips (neutron flux-and delta-T) and the reactor trip occurring in conjunction with receipt of the safety injection signal.
J 3.
Redundant isolation of main feedwater lines.
Sustained high feedwater flow would cause additional cooldown.
Therefore, in addition to the normal control action which will l
close the main feedwater valves, redundant safety injection sig-g nals will rapidly close all f eedwater control valves and f eed-water isolation valves.
4.
Trip of the f ast-acting main steam isolation valves (MSIVs) upon receipt of the following signals:
a.
Two out of three low steam line pressure signals in any one loop (above Permissive P-11).
(- I t
3
]
ENCLOSURE 1. TO NYN-87127 (Continued)
STEAM LINE BREAK CORE RESPONSE ANALYSIS ASSUMING.
CONSEQUENTIAL FAILURES DUE TO SUPERHEATED STEAM
{
b.
Two out of three High-2' containment pressure signals.
c.
Two out of three high negative steam line pressure I
l' rate signals in any one loop (below Permissive P-11).
The f ast-acting main-steam isolation valves are provided in each steam i
i line.
These valves are designed to close within 5 seconds upon receipt l
Q of an isolation signal. These valves are assumed to fully close within j
10' seconds following the occurrence of a large break in the steam line.
i For breaks downstream of the isolation valves, closure of all valves j
would completely terminate the blowdown'.
For any break, in any location, l
no more than one steam generator would experience an' uncontrolled blowdown even if one of the isolation valves fails to close unless a consequential.
f ailure results in reopening of both MSIVs in the pipe chase. This would I
cause an additional: cooldown of the primary system with a possible in-crease in the peak return to power. The following section of this report i
documents the analysis of these ef fects and consequences.
Besides the MSIVs, other equipment in the pipe chase that could be l
affected due to the presence of superheated steam are the steam gen-erator atmospheric relief valves- ( ARV) and the main feedwater isola-tion valves.
The ARVs are assumed to open as a result of a consequen-tial failure due.to superheated steam. The consequential reopening of the main feedwater isolation valves would not affect the transient since the safety injection signal also closes the feedwater control valves.
ANALYSIS OF EFFECTS AND CONSEQUENCES:
Method of Analysis i
The analysis of the steam line rupture, with consequential f ailures
[
due to superheated steam, has been performed to determines l
i 1.
The core heat flux and RCS temperature and pressure transients re-sulting from the cooldown following the steam line break described l
below. The LOFTRAN code modified for the calculation of super-heated steam enthalpy has been used.
This modified version util-izes the' basic LOFTRAN model and adds a calculation to determine the uncovered bundle region heat transfer (Reference 3).
2.
The thermal and hydraulic behavior of the core following the steam line break.
An evaluation of the plant parameters during this transient verifies that the Seabrook FSAR analysis is bounding, and that DNB, therefore, does not occur.
The analysis assumptions are the same as those of the steam line break f
analysis presented in Seabrook FSAR Section 15.1.5, with exceptions noted I
f '
0
ENCLOSURE 1 TO NYN-87127-(Continued)
STEAM LINE BREAK CORE RESPONSE ANALYSIS ASSUMING CONSEQUENTIAL FAILURES DUE TO SUPERHEATED STEAM i
herein. The most restrictive single failure in the Safety Injection System is assumed. The major difference between this analysis and the analysis presented in Seabrook FSAR Section 15.1.5 is that a consequential f ailure of affected equipment in the pipe chase due to superheated s team is assumed. - Specifically, the following scenario is analyzed:
1.
A 1.0 ft2 rupture occurs in one of the steam lines upstream' of the MSIV in one of the two pipe chases. The FSAR Chapter 15 analysis assumes a double-ended guillotine rupture of the main steam line with a break flow area of 1.4 f t2-(SG outlet nozzle) for forward flow and 0.8 ft2 (MSIV seat area) for reversed flow.
However, the. main steam lines in the pipe chase are "superpipe" and are therefore not subject to catastrophic failures such as a double-ended guillotine rugture as discussed in FSAR Subsec-tion 3.6(B).2,1.
The 1.0 ft rupture area is assumed for con-sideration of environmental ef fects in accordance with the re-quirement of Branch Technical Position ASB 3-1 attached to Sec-tion 3.6.1 of the Standard Review Plan (NUREG-0800). This case bounds smaller breaks because it results in a more severe cooldown l
of the reactor coolant system, and thus a greater peak heat' flux.
2.
All four steam generators blow down through the break until steam line isolation occurs by closure of the MSIVs.
3.
Initially, following steam line isolation, only the f aulted loop steam generator (loop 1) blows down.
Saturated s team is released to the pipe chase (shared by loops 1 and 2) until the tube bundle j
becomes uncovered, af ter which the steam exiting the break be-
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comes superheated.
j l
4.
Subsequently, the MSIVs and ARVs in the af fected pipe chase are
)
assumed to reopen when the tube bundle becomes uncovered.
The j
result is an uncontrolled blowdown of both affected steam gener-j ators (loops 1 and 2),
i To ensure conservatism in the analysis, cases were analyzed for both a
{
" late" and " predicted" cime for uncovering of the steam generator tube bundle.
In addition, cases were analyzed with reactivity feedback calculations based on the core properties associated with an assumed stuck'RCCA positioned in either the loop 1 or loop 2 core sectors.
RESULTS:
A time sequence of events for the four cases analyzed is shown in Table 1.
The consequential MSIV failures do not affect the peak return to power for cases in which reactivity feedback calculations are ' based on loop 1. since the peak occurs before the MSIVs open.
For
_4_
A l
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ENCLOSURE 1 TO NYN-87127 (Continued)
' STEAM LINE BREAK CORE RESPONSE ANALYSIS ASSUMING.
CONSEQUENTIAL FAILURES DUE TO SUPERHEATED STEAM the cases in which reactivity feedback is calculated based on loop 2, the! additional cooldown induced by the MSIV f ailures causes an in-creased' return to power.
Figures' I through 3 contain a set of plots showing the behavior of various core and plant parameters during the. transient of Case 1 (see. Table 1).
The transient of Case 3 is identical to Crae 1 un-til' the time of tube bundle uncovering and provides equally limiting results (i.e., the same minimum DNBR value).
However, only Case 1 results are provided since these ara' representative of both cases.
The results of Cases l' and 3 are more limiting than those of Cases 2 and 4 because the reactivity calculations are based on loop 1 (the faulted loop). Loop i experiences the most severe cooldown. Thus, weighting the reactivity feedback more on this loop maximizes the feedback, which provides a quicker return to criticality and a.
. greater peak return to power.
Since the peak return to power occurs prior to the consequential reopening of the MSIVs (see Table 1), the results are insensitive to the f act that two dif ferent assumptions for the uncovering of the tube bundle were made.
A. comparison of transient' parameters for all cases analyzed to the FSAR transient conditions verified that'the FSAR case bounds this analysis. Therefore, the DNBR remains above the limit 'value, and the conclusions of the Seabrook FSAR remain valid.
REFERENCES:
1.
PSNH Letter (SBN-1090),. dated June 5, 1986, " Steam Line Break Evaluation", J. DeVincentis to V. S. Noonan (NRC).
2.
Westinghouse Letter. dated March 25, 1986,-NS-NRC-86-3116,
" Westinghouse Response to Additional Request on WCAP-9226-P/
WCAP-9221-N-P, ' Reactor Core Response to Excessive Secondary Steam Release, (Non-Proprietary)".
3.
Osborne, M. P. and Love, D. S., " Mass and Energy Releases Following a Steam Line Rupture", September 1986, WCAP-8822-SI-P-A (Proprietary) and WCAP-8860-SI-A (Non-Proprietary).
4.
Safety Evaluation Report Related to the Operation of Seabrook Station, NUREG-0896, Supplement No. 5, July 1986.
f.
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l ENCLOSURE 1 TO NYN-87127 l~,
(Continued)
TABLE 1 l-TIME SEQUENCE OF EVENTS Tubes Uncovered'(Late),. Reactivity Calculations Based on Leop 1.
Case 1 Event Time (sec)
)
Steam line' ruptures -
0 j
Steam line-isolation occurs 9.8 Criticality attained 28.2 Peak heat flux reached 282.5 Steam generator tube bundle uncovered 412.2 Consequential f ailure of MSIVs and ARVs in 412.3 af fected pipe chase i
Tubes Uncovered (Late), Reactivity Calculations Based on Loop 2.
l Case 2
,1 Event Time (sec)
)
' Steam li ne, ruptures 0
Steam line isolation occurs 9.8.
Criticality attained 40.7 l
1 Steam generator tube bundle uncovered 478.9 j
Consequential f ailure of MSIVs and ARVs in 479.0 affected pipe chase 1
Peak Heat Flux reached 550.5 1
Case 3 - Tub 3s Uncovered (Predicted), Reactivity Calculations Based on Loop 1.
1 Event Time (sec)
Steam line ruptures 0
l Steam line isolacion occurs 9.8 1
Criticality att.ained 28.2 l
Peak heat flux reached 282.5 I
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I' 4
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ENCLOSURE l' TO NYN-87127
[
(Continued)-
' TABLE 1 i
LTIME SEQUENCE'0F EVENTS' l
P
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4 Steam generator - tube bundle uncovered 330.2-Consequential failure of MSIVs and ARVs in 330.3 L
affected pipe chase Case 4 '
Tubes Uncovered (Predicted), Reactivity Calculations Based on Loop 2.
Event Time (sec)
' Steam line ruptures 0
Steam line isolation occurs' 9.8 Criticality attained 40.7 Steam generator tube bundle uncovered 377.6 i
i Consequential f ailure of MSIVs and ARVs in 377.7 o
affected pipe chase Peak Heat Flux reached
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ENCLOSURE 2 TO NYN-87127' 1
REQUESTED CHANGES TO THE SEABROOK-TECHNICAL SPECIFICATIONS I
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3/4.1' REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUT 00WN MARGIN'- T,y GREATER THAN 20 W
' LIMITING CONDITION FOR OPERATION 3
3.1.1.1 The SHUT 00WN MARGIN for four-loop operation shall be greater than or equal to
- 1. 3% ak/k.h "^05 '.
l APPLICABILITY: MODES 1, 2*, 3, and 4.
ACTION:
With the SHUT 00WN MARGIN less than the limiting value, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required i
SHUT 00WN MARGIN is restored.
SURVEILLANCE _ REQUIREMENTS 4.1.1.1.1 The SHUT 00WN MARGIN shall be determined to be greater than or equal to the limiting value:
I a.
Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after detection of an inoperable control rod (s) and l
at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod (s) is inoperable.
If the inoperable control rod is immovable or untrippable, the above required SHUT 00WN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable 1
control rod (s)-
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b.
When in MODE 1 or H00E 2 with k,ff greater than or equal to 1 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6; When in MODE 2 with k,ff less than 1, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to c.
achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specifichion 3.1.3.6; i
d.
Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of Specifica-tion 4.1.1.1.le. below, with the control banks at the maximum inser-l tion limit of Specification 3.1.3.6; and i
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"See Special Test Exceptions Specification 3.10.1.
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