ML18086B487
| ML18086B487 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 04/19/1982 |
| From: | Vosbury F FRANKLIN INSTITUTE |
| To: | Nelson C NRC |
| Shared Package | |
| ML18086B486 | List: |
| References | |
| CON-NRC-03-81-130, CON-NRC-3-81-130 TER-C5506-82, TER-C5506-82-01, TER-C5506-82-1, NUDOCS 8205180045 | |
| Download: ML18086B487 (12) | |
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.1* TECHNICAL EVALUATION REPORT
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r PWR MODERATOR DILUTION ** J r PUBLIC SERVICE ELECTRIC AND GAS COMPANY r SALEM NUCLEAR GENERATING STATION UNIT 1
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.*. r* NRC DOCKET NO. 50-272 FRC PROJECT C5506
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NRCTACNO. 08398 FRC ASSIGNMENT 3 "*.. ~; .[-_ NRC CONTRACT NO. NRC-03-81-130 FRCTASK 82
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.*, r Prepared by Franklin Research Center Author: F. W. Vos bury r 20th and Race Street Philadelphia, PA 19103 FRC Group Leader: T. J. DelGaizo L Prepared for
.*J' [ Nuclear Regulatory Commission Washington, D.C. 20555 .Lead NRC Engineer: c. c. Nelsen
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April 19, 1982
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~j This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Goverhr.nent nor any agency thereof, or any-of their employees, makes any warranty, ex-
. pressed or implied; or asiumes any legal liability or responsibility for any
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l third party's use, or the results of such use, of any information, apparatus,
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1 product or process disclosed in this report, or* represents that its use by
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.1 TER-C5506-82 r
CONTENTS-Section .!!ill r l BACKGROUND 1 2 EVALUATION CRITERIA
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[ 3 TECHNICAL EVALUATION. 3 _*:1 r 3.l 3.2 Potential for Injection of NaOH Tank Contents Analysis of Most Limiting Moderator 3
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Dilution Incident 4. r 4 CONCLUSIONS
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5 REFERENCES
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TER-CSS06-82 j r FOREWORD
-": This Technical Evaluation Report.was prepared by Franklin Research Center r under a contract with the u.s. Nuclear Regulatory Conunission (Office of
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") assistance in support of NRC operating reactor licensing actions. The ~-J technical evaluation was conducted in accordance with criteria established by
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r the NRC. Mr. F. w. Vosbury and Mr. T. J. DelGaizo contributed to the technical r j
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- 1. BACKGROUND
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' r pressurized water reactor A limited moderator (boron) dilut.ion incident occurred at an operating (PWR) facility due to inadvertent injection of a r* portion of the NaOH tank contents into the reactor coolant system while the reactor was in a cold shutdown condition. Although only a small amount of the NaOH;solution (approximately 600 gallons) was injected and the reactor r remained subcritical by a large margin, the event highlighted the fact that a single failure.at this facility (misposition of an isolation valve) could l result in a previously unconsidered moderator dilution incident.
analysis, using certain extremely conservative assumptions, revealed that Subsequent
'" injection of the entire contents of the NaOH tank into the reactor coolant
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[' system could result in reactor criticality with the control rods inserted. Consequently, on September 19, 1977 [l], the NRC requested that Public
*i l Service Electric and Gas Company* (PSE&G) provide an analysis of the potential for and consequences of boron dilution accidents at the Salem Nuclear l Generating Station Unit l. The analysis was to be based upon the NRC's conservative assumptions consistent with the design and technical specifica-tions of Salem Unit l, including the assumption of the most limiting single failure. It also was to assess the factors affecting the capability of the
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1 operator to terminate postulated events prior to reactor criticality. PSE&G was requested to use the results of the analysis as the basis for Finally,
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*"l proposals for any corrective. action (design or procedural) required to preclude
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incidents. r\~.1 L PSE&G responded to theNRC's-request in a letter dated February l, 1978 '. -~
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?A l [2]. In reply to certain correspondence from the NRC,
- PSE&G submitted additional information concerning moderator dilution [3, 4, s, 6, 7). A conversation:[S)'
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-...i 2. EVALUATION CRITERIA fJ
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r The NRC provided dual criteria for this.review. determines that a moderator dilution incident could occur due to misposition First, when the Licensee
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r of the'isolation valve for the NaOH tank, proposed corrective action should be evaluated to determine if it significantly reduces the potential for such an incident. Any proposed changes are to be reviewed to ensure that they will
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I accomplish the intended purpose without adversely affecting plant engineered safety feature~ *
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r Second, where the Licensee determines that another potential moderator dilution incident is more limiting than the one analyzed in the FSAR, the r Licensee's analysis of the more limiting incident should be reviewed for
.1 acceptability in accordance with Section II of Standard Review Plan 15.4.6 *
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- 3. TECHNICAL EVALUATION 3.1 POTENTIAL FOR IN.Il:CTION OF NaOH TANK CONTENTS In Reference 2, PSE&G stated:
"The Salem Design consists of a 4,000 gallon sodium hydroxide tank which is connected to the recirculation line of each containment spray pump through an eductor. Pumping action is required to draw the contents of the sodium hydroxide*tank into the containment spray system. At Salem the only direct path to the RHR system would be from the containment spray pump recirculation line through the containment spray pump discharge
~ine, through valve CS36, and to the discharge line of the RHR pumps.
This path, however, is normally isolated when the RHR system is in the RCS cooldown modei when the path is in operation the direction of flow is from the RHR system to the containment spray system *
.~ f Based upon this review, we conclude that a Boron Dilution Incident at Salem is unlikely for the following reasons:
I 1. Two sets of shut-off valves exist in the potential flow path1 CS16 and CS17 in the sodium hydroxide tank discharge line and CS36 valves in the RHR-CS cross connect lines. Both sets of valves are in the closed position when the RHR system is in the RCS cooldown mode~
- 2. Test cycling, for surveillance requirements, of bOth sets of valves mentioned above is not simultaneous.
- 3. Pumping action is required to obtain considerable flow through an eductor. The containment spray pump recirculation lines are not normally operating during RHR system RCS cooldown mode.
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- 4. The RHR system, when not operating, is isolated from the containment spray system."
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<]t Evaluation A review of the Salem 1 FSAR [9] and system drawings confirmed that the 1 ( only flow path from the NaOH additive tank in the containment spray (CS) system into the reactor coolant system (RCS} was through the residual heat:' removal (RHR) system. When the RHR* system is in operation, two-valve protection from the NaOH tank is provided by shutting .the NaOH. tank discharge valves CS16 and CS17 and
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TER-C5506-82 the RHR-CS cross-connect valves llCS36 and l2CS36, as shown in Figure 1. In addition, a containment spray pump must be in operation and the CS pump discharge valves (llCS2. and i2CS2) must be *open in order* to transfer NaOH to the CS system. There is no possibility that unborated water from the NaOH tank would be injected into the RHR system during CS system testing or test cycling of either of the two sets of shutoff valves (CSl6, 17, or 36), since valves 11CS2 and l2CS2 are normally shut and the Licensee's procedures for testing CS
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Conclusion There is no potential for inadvertent boron dilution due to the mispos-itioning of either of the spray additive tank discharge valves (CS16 and
- 17) or the RHR-CS cross-connect valves (ll and l2CS36).
3.2 ANALYSIS OF MOST LIMITING MODERATOR.DILtJ'l'ION INCIDENT
.[ In Reference 7, PSE&G stated:
.:."J r *A thorough analysis was conducted on the Chemical and Volume Control System (CVCS) and all other interconnecting systems at all modes of reactor operation. Attention was directed towards identification of
- 1 possible paths for an inadvertent boron dilution of the Reactor Coolant System (RCS) to occur. Each path was analyzed as to the required modes of failure, if any, and the likelihood of occurrence.
TUbe failures of all heat exchangers located in the eves and other interconnecting (RHR, SI, etc.) systems was one area of evaluation. It was found that the Seal Water Exchanger has seal water return flowing at a lower pressure than that of the cooling water, component coaling. A postulated mode* of a failure for this heat exchanger was a single tube failure. Should this occur the total quantity of clean component cooling water leaking into the RCS would not cause a sharp drop in boron concentration, thereby initiating a sudden increase in ~eactivity. Tpe
- low level alarm in the component cooling sur*ge tank [volume 2000 gal] or high level of chromates in the RCS would notify the dperators of the problem. A total tube rupture was* considered to be extremely unlikely and was not *evaluated-. All other heat*exchangers are designed such that
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*e TER-C5506-82 the primary system pressure is greater than the cooling water system pressure, thus precluding the above situation from occurring.
A second possible path was primary water entering the eve system while flushing resins from the Ion Exchange Demineralizers. This process involves a total of 600-1,000 gallons of prim?ry water to be flushed with spent resins to the spent resin storage tank. The only possible path of entry of primary water into the eves would be due to a failure to close of *the process outlet valve located in the discharge line of each
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demineralizer **** In order to postulate the worst possible case it was J
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l r assumed that all 1,000 gallons enter the eves via the letdown line flowing to the Volume Control Tank (VCT). The amount of primary water flowing into the VCT depends upon the existing level in the tank. A three way'valve diverts letdown flow to the eves hold-up tanks on high
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[ level signals in the VCT. The portion of water flowing into the VCT enters as a spray mixing with approximately 1,000-2,000 gallons of
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borated water present in the tank. One charging pump normally takes
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- Total charging flow into the RCS runs as high as 100 gpm. This enters via the Reactor Coolant Pump Seals (20 gpm for all four pumps) and through the charging line to the RCS (55-80 gpm). Therefore, a situation
' [ could occur where there is 100 gpm of primary water entering the RCS. In order for this to occur, all 1,000 gallons of primary water must flow
-into the VCT with a minimum amount of mixing with the borated water l already present. The probability of this Oc:cutring is extremely low.
Nevertheless, if indeed the situation did arise in which 100 gpm of primary water was entering the RCS for a period of around 10 minutes and mixing with an approximate volume of 94,000 gallons of borated water the
- l. probability of an inadvertent boron dilution is minimal
- The reactor makeup portion of the eves was also reviewed. This area is
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This system is designed to limit the boron dilution rate such that under
*; various postulated failures, indication through instrumentation and
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;i one that is carried out under very strict administrative controls and
~:}~ l. also must adhere to the technical specification. Review of this portion of the eves did not uncover any postulated paths f~r inadvertent boron
~:J dilution that could not be corrected by operation action in a safe and **:J orderly manner.* *
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Evaluation The FSAR analysis on* uncontrolled :boron. dilution* was reviewed**: , The,FSAR' analyzed the case where unborated water was added to the RCS by both charging
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'l'ER-CS 50 6..;82 pumps at 300 gpm (RCS below operating pressure) and 236 gpm (RCS.at operating pressure).
During refueling (RCS volume 5717 cu ft), the boron concentration must be reduced from 2000 ppm to approximately 1500 ppm before the reactor would go r critical. This would take approximately 41 minutes and add 12,300 gallons of unborated water. action. A high source rate alarm would alert the operator to take r During startup (RCS volume 9967 cu ft), the minimum time required to r
- red~ce the boron concentration to where the reactor would go critical with all rods inserted would be 70 minutes (21,000 gallons). Again, the operator would be alerted by the high source rate count alarm and terminate dilution flow.
r During power operation, in automatic control at full po'wer, the control rods would reach the minimum limit of rod insertion in 7.2 minutes. Before f reaching this point, two alarms (low insertion limit and low-low insertion limit) would alert the operator to the uncontrolled dilution. [ If dilution continues after reaching the low-low alarm, approximately
; 14.3 additional minutes (3400 gallons) is required before the li shutdown *:ij l margin is lost. At full power, in manual control with no operator action, an overtemperature alarm and turbine runback, followed by a subsequent reactor l trip, would occur. The shutdown margin (li) will be lost in 14.3 minutes.
Therefore, the amount of water which would be added by the seal water heat exchanger tube rupture (2000 gallons) or by flushing of the ion exchange demineralizers (1000 gallons) is bounded by the uncontrolled boron dilution analysis in the FSAR. In all cases, the operator is provided with sufficient indications and alarms to allow him to take corrective action. Conclusion There are*no*potential boron dilution accidents which are not bounded by
*the current FSAR analysis *..
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- 4. CONCLUSIONS The following is a* *summary of conclusions regarding.* PSE&G' s review* of**
r potential moderator dilution incidents at Salem Unit l:
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r o There is no potential for inadvertent boron dilution due to the mispositioning of either the spray additive tank discharge valves (CS16 and 17) or the RHR-CS cross-connect valves (11 and l2CS36). o There are no potential boron dilution accidents which are not bounded
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REFERENCE:
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- l. G. Lear (NRC, . ORB) ..
. r Letter to F. P. Librizzi (PSE&G)
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Subject:
Boron Dilution Incident September 19, 1977
- 2. F. P. Librizzi (PSE&G)
Letter to G. Lear (NRC, ORB)
Subject:
Inadvertent Injection of NaOH February l, 1978 .. :.j
... 3. - s. A. Varga (NRC, ORB)
Letter to F. w. Schneider (PSE&G)
Subject:
Boron Dilution Accidents February 10, 1981
- 4. F. W. Schneider (PSE&G)
Letter to s. A. Varga (NRC, ORB)
Subject:
Boron Dilution Accident Analyses April 2, 1981
~- s. A. Varga (NRC, ORB)
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- l. Letter to F. w. Schneider (PSE&G)
Subject:
Request for Additional Information on PWR Moderator Dilution August 18, 1981 i I 6. F. W. Schneider (PSE&G) Letter to S. A. Varga (NRC, ORB)
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Subject:
Request for Additional Information on PWR Moderator Dilution October s, 1981
- 7. E. A. Liden (PSE&G)
'1 L l '>1 Letter to S. A. Varga (NRC, ORB)
Subject:
PWR Moderator Dilution.
*.:J February 3, 1982
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*.,*j 8. M. Gibbons (PSE&G) .. ,*" ..
. ~~ C. Nelson (NRC)
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*1 I Telephone conversation with F. Vosbury (FRC)
April 6, 1982
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- 9. Salem Nuclear Generating Station Units l and 2
-,," l Final Safety Analysis Report, through Amendment 29 Public Service :Electric.and Gas* Company ...... *:;. .<.*.. :*
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