ML20209D658
| ML20209D658 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 07/02/1999 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20209D648 | List: |
| References | |
| NUDOCS 9907130192 | |
| Download: ML20209D658 (6) | |
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d UNITED STATES 3
E NUCLEAR REGULA1 ORY COMMISSION If WASHINGTON, D.C. 20555 000t
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION BELATED TO AMENDMENT NO, 172 4
TO FACILITY OPERATING LICENSE NO. NPF-49 NORTHEAST NUCLEAR ENERGY COMPANY. ET AL.
MILLSTONE NUCLEAR POWER STATION. UNIT NO. 3 POCKET NO. 50-425
1.0 INTRODUCTION
By letter dated June 5,1998, as supplemented by a letter dated sanuary 13,1999, the Northeast Nuclear Energy Company, et al. (the licensee), submitted a request for changes to the Millstone Nuclear Power Station, Unit No. 3 (MP3), licensing basis. The requested changes would revise the MP3 licensing basis associated with the design basis steam generator tube rupture (SGTR) accident analysis described in Chapter 15 of the MP3 Final Safety Analysis Report (FSAR) to address an unreviewed safety question. The SGTR analyses described in the FSAR include an offsite dose analysis and a margin to overfill analysis. Both of these ana;yses were updated. The offsite dose analysis was updated to reflect a larger capacity for the steam generator atmospheric dump valve (ADV) and a decreased operator response time to close the ADV block valve, and the margin to overfill analysis was updated to reflect a new single failure on the atmospheric dump bypass valves (ADBV) that could cause failure to open on demand of the ADBV associated with two of the intact steam generators. The revised licenaing basis will be incorporated into the FSAR and will revise the SGTR accident analysis to address the changes in the offsite dose and margin to overfill analyses. The [[letter::B17490, Forwards Response to NRC 980910 & 1218 RAIs Re Util Proposed LAR Re Revised SG Tube Rupture Analysis, .No New Regulatory Commitments Are Contained within Ltr|January 13,1999, letter]] provided clar;fying information that did not change the initial proposed no significant hazards consideration determination.
2.0 EVALUATION The licensee's revised SGTR analysis uses the same methodology as the current analysis.
This methodology i.s described in WCAP-10698 'SG TR Analysis Methodology to Determine the Margin to Steam Generator Overfill," March 30,1987, that was approved by the staff. The revised analysis for radiological consequences assumes a larger ADV flow capacity of 820,000 lb/hr/ valve. However, the operator action time for the closure of the block valve to isolate the stuck-open ADV on the ruptured steam generator is assumed to be,20 minutes in the revised analysis instead of 30 minutes as assumed in the current analysis. The quicker closure of the
- block valve more than compensates for the larger capacity assumed for the ADV. The results of the r;evised analysis show that the radiological consequences are not increased from those previously calculated.
9907130192 99070.2 PDR ADOCK 0S000423 P
. The revised analysis of a margin to steem generator (SG) overfill assumes a moot IMiting single failure that causes failure to open of the ADBV cssociated with two of the intact steam generators (SGs) when the operator initiates cooling of the reactor coolant system using the intact SGs. In this part of the analysis, a larger minimum capacity for the ADBV of 820,000 Ib/hr/ valve is credited. The licensee states that this larger flow capacity is still a conservative minimum capacity for the ADBV. The operator actions and their limiting time credited in the SG overfill analysis are not being changed. The results of the licensee's revised analysis shows that sufficient margin still exists for SG overfill following a design basis SGTR event.
Operator Response Times The staff's evaluation of WCAP 10698 stipulates plarit-spscific criteria for assessing operator 1
response times in the event of an SGTR. The staff used those criteria to evaluate the
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information submitted by the licensee regarding operator response times during an SGTR at j
MP3. The staff based lis sWtion on the June 5,1998, submittal, as supolemented in the l
[[letter::B17490, Forwards Response to NRC 980910 & 1218 RAIs Re Util Proposed LAR Re Revised SG Tube Rupture Analysis, .No New Regulatory Commitments Are Contained within Ltr|January 13,1999, letter]].
Criterion 1:
Provide simulator and emergency operating procedure (EOP) training
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related ts a potential SGTR.
The licensee documented by [[letter::B17490, Forwards Response to NRC 980910 & 1218 RAIs Re Util Proposed LAR Re Revised SG Tube Rupture Analysis, .No New Regulatory Commitments Are Contained within Ltr|letter dated January 13,1999]], that onsite simulator and EOP
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training relevant to an SGTR are provided. The staff finds that the licensee has satisfied Criterion 1.
l Criterion 2: U'llizing typical control room staff, complete demonstration runs to show that the postulated SGTR accident can be mitigated within a period of time compatible with overfill prevention.
l By [[letter::B17490, Forwards Response to NRC 980910 & 1218 RAIs Re Util Proposed LAR Re Revised SG Tube Rupture Analysis, .No New Regulatory Commitments Are Contained within Ltr|letter dated January 13,1999]], the licensee submitted the demonstrated operator response i
tir.ws for the overfill scenario. The demonstrated response times, compared with the times I
assumed in the licensee's SGTR analysis, are given in the following table:
Operator Action Demonstrated Time
- Assumed Time (minutes: seconds)
(minutes)
Event initiation (malfunction irserted) to feed flow 17:12 16.5 to rupture-d SG stopped (E-3, step 4.b)
Feed flow to ruptured SG stopped (E-3, step 4.b)so 6:21 8.0 reactor coolant system (RCS) cooldown initiated (E-3, step 14.b)
RCS cooldown terminated (E-3, step 14.d) to RCS 1:07 3.0 depressurization initiated (E-3, step 18.b) l RCS depressurization terminated (E-3, step 18.e) 1:29 3.0 to all but one charging pump stopped (E-3, step 21) l Total 26:09 30.5 "The demonstrated response time was derived from the arithmetic mean of response times from all crews, both operations and administ;ation c:ews.
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I-Further, the licensee indicated that simulation runs were completed with three administration
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- crews and five operations crews.'The staff derived the arithmetic mean for operator response times, which were used as the demonstrated times (shown in th'e preceding table) for the i
overfill scenario. The licensee stated in its January 13,1999, submittal that one operations crew failed to meet the total assumed time of 30.5 minutes. The licensee also stated that during retesting this crew was successful on two different occasions, with total demonstrated times of 24 and 20 minutes. The average demonstrated operator response time (17 minutes and 12 seconds) for " feed flow to ruptured SG stopped" was 42 seconds greater than the time assumed (16 minutes and 30 seconds) in the licensee's SGTR analysis. The staff found this y
discrepancy to be acceptable because the average total operator response time (26 minutes and 9 seconds was bounded by the total assumed time (30 minutes and 30 seconds).
In addition, the licensee's January 13,1999, submittal presented information on the control room staff's ability to (1) determine that the ruptured SG's ADV had failed'open, and (2)
- complete actions to close the block valve of the ruptured steam generator's ADV within the required time of 20 minutes. The licensee stated that each ADV has indicating lights that show whether the valve is open or closed. The licensee also stated that there is an annunciator on Main Board 5, " Main Steam Relief Valve Not Closed,' that annunciates when an ADV opens.
- To close the block valve of the ruptured SG's ADV, the operator must push the "close" button
' and check the position indication to verify that the valve closed.' Finally,,the licensee stated that
. during informal sessions two crews closed the subject block valves in less than 10 minutes.
Crews will be formally timed in the 1999/2000 licensed operator requalification training cycle.
- The staff finds this information to be acceptable.
. On the basis of the information discussed, the staff finds that the licensee has satisfied Criterion 2. The staff also finds that the licensee has given satisfactory assurance that
- operators can identify'that the ruptured SG's ADV is open, and then can close the associated block valve within the required time of 20 minutes.
Criterion 3: If the emergency operating procedures (EOPs) specify SG sampling as a means of identifying the SG with the ruptured tube, provide the expected time period for obtaining the sample results and discuss the effect ori t%
duration of the accident.
By [[letter::B17490, Forwards Response to NRC 980910 & 1218 RAIs Re Util Proposed LAR Re Revised SG Tube Rupture Analysis, .No New Regulatory Commitments Are Contained within Ltr|letter dated January 13,1999]], the licensee stated that SG sampling is performed daily. The licensee noted that counting the sample and identifyir'g which generator has a ruptured tube would take about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The licensee explained tha the leak rate can be estimated by sampling the condensate polishing facility effluent and steam Jet air ejector..The licensee
,. indicated that this sampling can be performed concurrent with the SG sampling and is also estimated to take about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The licensee stated that SG tube leakage so small that the l
sampling that would be required to identify the SG with the ruptured the would be bounded by'
-the analysis performed for the complete severance of one SG tube for both the overfill and -
dose effects. In addition, the licensee stated that if an unexpected increase in SG level has not occurred and sampling is required to identify the SG with the ruptured tube, then steam generator overfillis not a concem. On the basis of this information, the staff finds that Criterien 3 is satisfied.
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l Dose Assessment The staff reviewed the inputs and assumptions of the licensee's offsite dose calculation and found the licensee's analysis to be acceptable. The staff performed a confirmatory analysis using the licensee's assumptions for the SGTR and calculated comparable results. Staff analysis assumptions and results are presented below, along with the licensee's results. Staff dose calculations also assumed the 20-minute closure time for the affected SG ADV block valves proposed by the licensee and found to be acceptable by the staff as previously discussed. The licensee's calculated offsite dose consequence results meet the acceptance criteria in 10 CFR 100, and thus are acceptable. The current FSAR control room dose is evaluated for the limiting loss-of-coolant accident (LOCA). The requested changes to the SGTR dose analysis offset one another such that the calculated offsite doses decrease. For this case, the staff concludes that the postulated dose to operators in the control room would likely also decrease. Therefore, the staff did not evaluate the control room dose due to the revised SGTR analysis. The staff finds the proposed change to be acceptable with regard to the radiological consequences of the SGTR.
Offsite Radiation Dose Results:
Dose to Thyroid (rem)
Licensee Staff 10 CFR 100 Calculated Calculated Acceptance Criteria
- 1. Accident initiated lodine Spike Exclusion Area Boundary 19 18 30 Low Population 7' e 2.0 1.4 30 i
- 2. Pre-accident lodino Spike Exclusion Area Boundary 51 50 300 Low Population Zone 4.0 3.6 300 Staff Assumptions for SGTR Analysis:
Iodine species 100% elomontal Reactor Coolant lodine Activity Accident initiated Spike 1.0 pCi/gm of Dose Equivalent (D.E.) 1-131, with iodine spike 500 times larger at assumed appearance rates given (see Tables 1 and 2 below)
Preaccident Spike 60 pCi/gm of D.E.1-131 (see Tablo 1)
Secondary System initial Activity 0.1 mci /gm of D.E.1131 (see Table 1) m
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Table 1 - lodine Activities in Primary and Secondary Coolant Activity (Ci)
Primary Coolant D.E.1-131 Secondary Coolant D.E.1-131
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Nuclide 1.0 uCi/am 60 uCi/am 0.1 uCi/am (per SG) 1-131 108 10804 3.36 l132 63.5 3798 1.12 1-133 283 16985 4.79 l-134 39.6 2383 0.204 l-135 152 9153 2.10 j
Table 2 - lodine Spike Accearance Rates Nuclide Ci/sec Ci/hr I-131 1.70 6120 l-132 3.24 11664 l-133 3.84 13680 1-134 4.71 16956 l
l-135 3.59 12924 Mass water in RCS (Ibm) 520,000 Mass water initially in each SG (Ibm) 95,170 Break Flow (0-2 hrs) (Ibm) 211,400 Loss of offsite power at time of reactor trip (sec) 109 Faulted SG ADV Block valve closed (min) 20 Primary-to-secondary leakage (hrs) 8 Total primary-to-secondary leakage (gpm) 1.0 Atmospheric Dispersion Factors EAB X/O (s/m )
0 - 2 hrs 4.3 E-04 LPZ X/O (s/m )
0 - 8 hrs 2.9 E-05 Activity release data from submittal Table 15.6.3-3, Figure 15.6.3-6.
and Figure 15.6.3-12 lodine partition coefficient 0.01
3.0 STATE CONSULTATION
in accordance with the Commission's regulations, the Connecticut State official was notified of the proposed issuance of the amendment. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of ar y effluents that may be released offsite, and that there is L
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. no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (63 FR 35992). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
5.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: Michelle Hart C. Y. Liang j
Garmon West
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Date:
July 2,1999 o
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