ML18092B168
| ML18092B168 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 06/04/1986 |
| From: | Corbin McNeil Public Service Enterprise Group |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| NLR-N86084, NUDOCS 8606100191 | |
| Download: ML18092B168 (7) | |
Text
Public Service Electric and Gas Company Corbin A. McNeill, Jr.
Public Service Electric and Gas Company P.O. Box236, Hancocks Bridge, NJ 08038 609 339-4800 Vice President -
Nuclear June 4, 1986 NLR-N86084 Off ice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission 7920 'Norfolk Avenue Bethesda, MD 20014 Attention:
Mr. Steven A. Varga, Director PWR Project Directorate #3 Division of PWR Licensing A
Dear Mr. Varga:
SERVICE WATER HEADER OUTAGE REVISIONS TO SAFETY EVALUATION SALEM GENERATING STATION UNIT NOS. 1 AND 2 DOCKET NOS. 50-272 AND 50-311 In letters, dated August 30, 1985 and December 19, 1985, PSE&G requested amendments to Facility Operating Licenses DPR-70 and DPR-75 for Salem Generating Station, Unit Nos. 1 and 2.
These amendments were to permit operation with one service water header out of service in modes 5 and 6.
On March 7, 1986, the Commission approved the request and issued Amendment No. 72 and No. 46 to Facility Operating Licenses DPR-70 and DPR-75, respectively.
The safety evaluation submitted in support of the amendment request made certain assumptions regarding the time interval between reactor shutdown and Reactor Coolant System (RCS) drain down.
However, as a result of procedural changes implemented during the recently completed Salem Unit 1 refueling outage, it was determined that the time interval could be reduced from the originally specified four days to as little as two days.
In order to utilize a shorter time interval, the assumptions made in the safety evaluation were reviewed for necessary changes.
An evaluation of the impact of each change on the conclusions of the safety evaluation was then performed.
It was determined that the changes did not alter the conclusions of the original safety evaluation and, therefore; RCS drain down at two days after reactor shutdown is acceptable.
8606100191 860604 PDR ADOCK OS000272 p
Mr. Steven 6-4-86 We are forwarding a copy of the evaluation described above for your review.
The proposed change and the results of the evaluation have been discussed with Mr. D. Fischer and Mr. w. Jenson of your staff.
Should you have any questions regarding this matter, please feel free to contact us.
Attachment C
Mr. Donald c. Fischer Licensing Project Manager Mr. Thomas J. Kenny Senior Resident Inspector Sincerely,
INTRODUCTION IMPACT OF CHANGES IN ASSUMPTIONS RELATIVE.TO LCR es~1e On March 7, 1986, the NRC approved a change to the Salem Units 1 and 2 Technical Specifications (Amendments 72 and 46, respectively) which allowed a service water header to be inoperable for inspection or maintenance while the plant is in modes 5 or 6.
As a result of experience gained during the recently completed Salem Unit 1 refueling outage, PSE&G wishes to revise a few of the assumptions used in the safety evaluation which was submitted in support.of the change to the Technical Specifications. Specifically, these revisions would allow the originally specified Reactor Coolant System (RCS) drain down time to be changed from four (4) days to two (2) days after reactor shutdown.
The purpose of this evaluation is to document the revisions and demonstrate that those revisions continue to support a conclusion that the change to the Technical Specifications did not involve a significant hazards consideration.
The current wording of the Technical Specifications is not impacted by the revisions discussed in the following sections.
A.
EARLIER RCS DRAIN DOWN TIME The first assumption change deals with the time interval between
- reactor shutdown and RCS drain down in mode 5.
The original evaluation assumed that the RCS would not be drained down to the reactor vessel nozzle centerline until four days after plant shutdown, which was thought to be the minimum time at which drain down could occur.
However, new procedures for degassing and cleaning up the reactor coolant system have been developed which can result in the RCS being ready for drain down as early as two days after shutdown.
The main effect of the earlier drain down is that a higher decay heat load will initially be present with the RCS drained down to the nozzle centerline and one service water header out of service.
A summary of the areas in the original safety evaluation impacted by this change along with an evaluation of the impact is given below.
- 1.
ORIGINAL EVALUATION The minimum core uncovery time, if RHR is lost, is given as 1-1/2 hours (90 minutes) at four days after shutdown (Section 4.3.1)
- IMPACT Instantaneous RCS drain down at the two days after shutdown reduces the minimum core uncovery time to approximately 72 minutes.
- - EVALUATION OF IMPACT The intent of calculating core uncovery time was to demonstrate that, if normal residual heat removal is lost, sufficient time is available for operator action to line-up back-up decay heat removal (DHR) flowpaths.
Since the core uncovery time is itill greater than one hour, the operator still has sufficient time to take the required action (a detailed procedure has been developed which details exactly what action needs to be taken if all service water is lost)
- Therefore, the intent of the original evaluation is still met.
- 2.
ORIGINAL EYALUATION The time available for decay heat removal when using the short-term alternate heat removal flowpaths identified (use of RWST and spent fuel pits as heat sinks) was stated to range from two to nine hours, during the initial ten day period when the water level is at the reactor vessel nozzle centerline and the RWST and spent fuel pits are initially at maximum temperatures of 98°F.
Based on a more realistic initial temperature of 80 F, the minimum DHR time was given as five hours (Section 4.3.2).
IMPACT When the RCS is drained down to the reactor vessel nozzle centerline at two days after shutdown, the time available for DHR using these short-term alternate heat removal f lowpaths will vary from 1-1/2 to 8-1/2 hours based on initial temperatures of 9a°F.
Based on the more realistic initial temperature of 80 F, the minimum DHR time will be greater than three hours.
(In reality, the initial temperature of the RWST is typically even less than 80°F during the time when refueling outages occur so that the DHR time will likely be longer)
- EVALUATION OF IMPACT As stated in the original safety evaluation, this short-term heat removal process is not required to sustain a safe shutdown condition but is employed for economic considerations to avoid having to initiate the make-up/boil-off process, if all service water is lost.
The intent of presenting this alternate heat removal path in the original evaluation was to demonstrate that significant time existed to correct any credible problems and, if necessary, to line-up the make-up and boil-off heat removal process for long term decay heat removal.
Since significant time still exists for decay heat removal using the RWST and spent fuel pits as heat sinks, the intent of the original evaluation is still met.
- e.
- 3.
ORIGINAL EVALUATION It was stated that the core uncovery time, if RHR is lost, can be increased to six hours if the RCS is filled up to the reactor vessel flange elevation (Section 4.3.2).
IMPACT At two days after shutdown, the minimum core uncovery time with the RCS filled to the reactor vessel flange, is reduced to approximately 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
EVALUATION OF IMPACT The intent of this section was to demonstrate that the core uncovery time could be further increased, allowing even more time for correcting any problems or lining up the short term heat removal process, via the RWST/SFP's.
A significant increase in core uncovery time (4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) can still be achieved by filling the RCS to the reactor vessel flange.
Therefore, the intent of the original evaluation is still met.
- 4.
ORIGINAL EVALUATION It was stated that a f lowrate of approximately 80 gpm would be required to maintain constant level at the nozzle centerline if the make-up/boil-off process is utilized (Section 4.3.3)
- IMPACT At two days after shutdown, the required flowrate increases to approximately 90 gpm.
EVALUATION OF IMPACT Because the 90 gpm flowrate can easily be supplied by either of the two safety grade pumps (any combination of charging, safety injection or containment spray pumps) that are maintained available, this change has no effect.
- 5.
ORIGINAL EVALUATION In the PSE&G response to Questions 6 and 7 of NRC "Request for Additional Information on the Service Water Header Outage", dated December 19, 1985, results of dose analyses were presented for the heat removal process utilizing the RWST as a heat sink (Question 6) and for the make-up and boil-off process (Question 7)
- The source term used for these analyses was based on the activity available at four days after shutdown.
IMPACT The dose analyses were re-evaluated based on a source term corresponding to two days after shutdown.
As a result, the source terms for noble gas and iodine increase by 39% and 56%, respectively.
This yields a corresponding increase of 39% and 56% in whole body and thyroid doses.
EVALUATION OF IMPACT As can be seen from the results presented in response to questions 6 and 7, a 39% and 56% increase in whole body and thyroid doses would still be significantly below the 10CFR100 allowable limits.
Therefore, this impact is acceptable.
In conclusion, initiating RCS drain down at two days after shutdown results in some decrease in the minimum core uncovery time and decay heat removal time available with the short-term alternate heat removal flowpaths.
However, sufficient time still exists for the operator to take the appropriate action, if normal RHR is lost, to either restore the normal RHR flowpath, line-up the alternate short-term DHR flowpaths or line-up the long term safety-grade back-up DHR flowpath (make-up/boil-off process)
- The increased radiation exposures due to the change are acceptable as the calculated doses at the site boundary are still significantly below the 10CFR100 allowable limits.
B.
ALLOWABLE SPENT FUEL PIT TEMPERATURE RISE The second assumption change deals with the allowable temperature rise in the spent fuel pools if the alternate short-term DHR process is employed.
The original safety evaluation (Section 4.3.2) stated that th5 spent fuel pools would only be permitted to heat up to 100 F.
This was based on a conservative arbitrary upper limit for the pool temperature.
Subsequently, it was decbded to increase this allowable temperature limit to 120 F, which is consistent with the design limit of the pools.
This change increases the DHR time available when using the fuel pools as heat sinks and further minimizes the chances of having to initiate the make-up/boil-off process.
Thus, this change has a net positive effect on the evaluation.
In addition to the two assumption changes discussed above, one area of discussion relative to this technical specification needs to be clarified.
In the PSE&G response to an NRC "Request for Additional Information on the Service Water Header Outage",
dated December 19, 1985, it was stated (Question 4) that all operators will be trained on the new emergency procedures for loss of all service water.
This was intended to mean that all licensed operators will be trained on the new procedures.
Consistent with other emergency operator procedures, only licensed operators are required to be trained on the procedures.
It should be noted that this technical specification change was successfully implemented during the recent Salem Unit 1 refueling outage.
Detailed procedures have been written which identify the steps that must be taken prior to entering the desired configuration (one service water header out of service with the RCS partially drained down).
Detailed emergency procedures have been written that identify steps to be taken if all service water should be lost. All licensed operators were trained on both of these procedures prior to the refueling outage.
All the required temporary hoses and portable fans were in place to support operation of various pumps if service water were lost.
In addition, a method of measuring reactor vessel level using all safety grade instrumentation with battery back-up was developed and successfully implemented.
This represents an additional improvement over the level measurement system discussed with the NRC which did not have battery back-up.