ML19331D167
| ML19331D167 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 08/25/1980 |
| From: | Hoffman D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8008270401 | |
| Download: ML19331D167 (5) | |
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'h General Offices: 212 West Michigan Avenue. Jackson Micnigan 49201. A rea Code 517788-0550 August 25, 1980 Director, Nuclear Reactor Regulation Att Mr Dennis M Crutchfield, Chief Operating Reactors Branch No 5 U S Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - PROPOSED PROBABILISTIC RISK ASSESSMENT:
ADDITIONAL INFORMATION FOR NRC REVIEW Consumers Power Company letter dated February 22, 1980, proposed that a probabil-istic risk assessment of Big Rock Point be performed. Deferral of certain plant modifications was requested pending completion of this assessment. Additional information regarding modifications for which deferral was requested was provided by letter dated April 2, 1980.
NRC per::ennel visited Big Rock Point on July 28, 1980, as part of their review of our February 22, 1980, request.
During this visit, various agreements were reached and commitments were made in the areas of emergency diesel generator testing and adequacy of the interim Technical Support Center. These agreements and commitments are documented in the attachments to this letter.
David P Hoffman (Signed) l l
l David P Hoffman i
Nuclear Litensing Administrator CC Directcr, Region III, USNRC NRC Resident Inspector - Big Rock Point Attachments h0]
80 08 270
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INTERIM TECHNICAL SUPPORT CENTER EVALUATION BIG ROCK POINT The interim Technical Support Center (TSC) at Big Rock Point was evaluated during an NRC visit on July 28, 1980.
This evaluation was a part of NRC review of Consumers Power Company's February 22, 1980, request to defer construction of a permanent center pending completion of a probabilistic risk assessment.
During the evaluation, additional commitments were made in certain areas.
These commitmentsand the areas discussed during the visit are described below.
Technical Support Center Manning:
The requirements for manning the TSC, including the personnel designated to do so, were reviewed.
It was agreed that appropriate expertise would be provided.
Concern was expressed, however, over whether personnel would be capable of return-ing to the plant to man the TSC in the event of an accident which released sig-nificant amounts of radioactivity to the containment atmosphere. This concern results from the predicted high radiation levels which might be encountered out-side the unshielded steel containment sphere.
Consumers Power Company agreed to further evaluate the radiation exposures which might be received during plant ingress and identify any conditions in which TSC manning might be prohibited.
It is not possible to quantitatively evaluate specific accident sequences which might preclude TSC activation at this time.
This is because no previously analyzed accident sequence has predicted release of a significant amount of the core inven-tory of radioactivity. Such an accident would have to involve a significant degradation of core cooling. Emergency core cooling at Big Rock Point is provided by an automatic reactor depressurization system (RDS) and two low pressure core spray systems. The BDS consists of four trains of depressurzing valves and associ-ated logic, each povered by its own uninterruptible power supply. Proper actuation of three of four trains is sufficient to depressurize the plant and permit core spray activation before significant core damage occurs.
A situation like that which occurred at Three Mile Island, a partially uncovered core with the plant still pressurzied for an extended period, would require a complete failure of RDS.
Each of the low pressure core sprays has been tested in a full scale mockup with steam back pressures representative of those predicted during a Big Rock Point depressurization. These tests demonstrated that either core spray, acting indi-vidually, can provide sufficient spray flow appropriately distributed to complete-ly remove core decay heat. Significant degradation of both core sprays, like complete failure of RDS, has not been previously analyzed. Accident sequences such as these vill, of course, be considered in the proposed risk assessment.
In the absence of specific accident sequences which have been predicted to cause significant core damage, our evaluation of plant ingress radiation exposures has considered the core inventory release fractions specified in NUREJ-0578. This assumes instantaneous release of 100% of the noble gases and 25% of the halegens, into the contain=ent atmosphere at the time of the accident.
(Radioactivity released to the water in the lower portions of the containment does not contribute significantly to site radiation levels since the post-accident water level remains below grade level). The times required for plant ingress were verified by timing personnel traversing these routes. The fractions of total core inventory release (at the above specified fractional constituency) which could be acconmodated while limiting ingress exposures to specific levels then vere evaluated.
2 The evaluation identified that individualingress exposures can be limited to three rem if approximately k0% of the core fission product inventory is released. This assumes that the ingress occurs one-half hour after the accident.
If individual exposures are permitted to approach 25 rem (as recommended by paragraph 259 of National Council :en Radiation Protection and Measurements Report No 39, " Basic Radiation Protection Criteria" for emergency conditions where protection of facili-ties or elimination of further escape of effluents is necessary) plant ingress to man the TSC would be possible even for a complete release of the core fission product inventory.
In this case, individualingress exposures would be approximately T.2 rem.
As a result of the evaluation described above, Consumers Power Company is assured of ability to man the TSC under any accident conditions.
Procedures to Estimate Core Damage:
Interim procedures to quantify core damage in the event of an accident (as discuss-ed in CPCo letter dated December 27, 1979, January 18, 1980, and March lb, 1980) were reviewed.
Copies of Energency Plan Implementing Procedures 5B, " Procedure to Determine High Stack Gas Releases" and 5D, " Procedure to Determine Extent of Core Damage (for 0% to 100% Core Meldown)" were provided to the NBC reviewers and were discussed. The reviewers indicated that the methods and procedures for accomplish-ing these estimations were acceptable.
Availability of Plant Status Information:
The NRC reviewers evaluated the ability to obtain plant status information in the TS C.
It was agreed that much information could be ascertained by observation through the control room vindows due to the configuration and small size of the control room.
It was agreed that an individual should be designated whose princi-ple responsibility is to assure adequate information transfer and that this indi-vidual should be empowered to enter the control room and obtain data if necessary; emergency procedures vill be revised to so designate one or two TSC personnel.
In addition, status boards will be mounted in the TSC to permit manual display of critical plant parameters in clear view of TSC personnel.
Control Room /TSC Habitability:
The potential radiation and radioactivity levels in the control room /TSC area following an accident involving major core da= age were discussed.
It was agreed that the shielded area, although small, was adequate to assure ability to control operations in the initial post accident period when radiation levels are of great-est concern.
Protection of personnel from airborne radioactivity which might be present in the event of containment leakage and unfavorable atmospheric dispersion can only be acecmplished using self-contained breathing apparatus.
Consumers Power Company vill procure, as quickly as.possible, additional equipment as required to ensure that a self-contained breathing unit is available in the control room, TSC, or operations support center for each member of the normal shift oper-ating staff (six personnel) and for each person designated in the e=ergency preced-ures for i==ediate call back to man the TSC.
Five extra units vill also be provid-ed.
Refill / replenishment capability sufficient for a six hour period will be pro-vided. A six hour capability will envelope the period between a design basis accident and the time at which contain=ent pressure vill return to essentially atmos-pheric, thereby reducing the potential for leakage.
EMERGENCY DIESEL GENERATOR TESTING BlG ROCK POINT NRC Regulatory Guide 1.108 addresses requirements for periodic testing of emergency diesel generator (EDG) units. The Regulatory Guide includes recommendations appli-cable to design of EDGs to facilitate testing, preoperational testing, periodic testing, and acceptable frequencies of periodic testing. Big Rock Point's EDG predates the Regulatory Guide; the recommendations applicable to design and preoper-ational testing therefore cannot be implemented.
Consumers Power Company does intend to meet the intent of the reco=mendations concerning periodic testing (including frequency) and record keeping.
Design differences and differences between past test-ing and the Regulatory Guide recommendations make it impossible to implement these sections of the Guide directly; accordingly, the testing program which will be followed is described below.
Test Frequency:
The Regulatory Guide (Section 2.d) recommends a variable testing frequency based on actual failure rates. The Guide specifies that only " valid" tests as specified in Section 2.e be considered in establishing the testing frequency.
Consumers Power Company's past testing practice has been to periodically load the EDG using the electric fire pump (51% of maximum expected EDG loading) for 20 to 30 minutes.
These tests would not be considered " valid" per Section 2.e since that section requires a loaded period of at least one hour. Without including these tests, however, sufficient failure rate data does not exist to establish a test frequency. Accordingly, all tests in which the EDG was operated under load are considered " valid" tests for purposes of establishing an initial test frequency.
Test data has been assembled for the period since January 1, '978.
During this period, there have been 65 " valid" tests. Five tests were declared failures.
Two of these failures, however, only involved failure to start within a very short acceptance time; the acceptance time has been extended (from 13 9 seconds to 31.2 seconds) based on a recent, more detailed, evaluation of the time sequence associated with a loss of coolant accident coincident with loss of offsite power,
=aking these two tests acceptable. The remaining three failures in 65 tests yield a failure rate of h.6%.
The Guide dictates a test interval of once per three days for failure rates greater than h%. Testing at this interval was initiated August 18, 1980, and will continue until failure rate data permits a revised interval in accord-ance with the Guide.
Test Procedure:
Section 2.c of the Guide specifies recc=mendations for the conduct of periodic testing. These specify that tests must de=onstrate proper starting and verify that required frequency and voltage are attained within acceptable limits and time.
Co=ponents required for automatic startup are to be verified operable.
" Full-load-carrying" capability is to be demonstrated for at least one hour such as by synchronizing the generator with offsite power and assuming a load at the
=aximum practical rate. The cooling system =ust function within design limits.
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Emergency Diesel Generator Testin8 2
Big Rock Point Plant Periodic testing of the Big Rock Point EDG will involve opening a Bus Tie Break-er to verify proper operation of the automatic start system. The EDG vill be verified to start within 31.2 seconds with output frequence of 60f,2 E: and output voltage of h80V+10%. There are no provisions fev synchronizing the EDG vith offsite power; the maximum practical load is the electric fire pump (51% of maxi-mus expected load). Addition of other loads during EDG testing without being able to synchronize with offsite power would affect availability / operability of equip-ment important to safety.
Thus, the fire pump will be used as the EDG load, and the EDG will be run under load for at least one hour. (The EDG is alsc tested using a resistive load bank to 95% of rated capacity during each refueling outage.) The cooling system vill be verified to operate within design li=1ts.
Valid Tests and Failures:
Section 2.c of the Guide specifies criteria for valid tests and failures. These criteria were used without modification during the review of historical data described above, except that loaded tests of less than one hour duration were considered valid. The test procedure has been modified to require loaded oper-ation for at least one hour. EDG tests occurring on or after August 18, 1980,
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will be classified as " valid" and/or " failures" in accordance with the criteria presented in Section 2.e.
Records and Reports:
Section 3 of the Guide specifies record keeping and reporting requirements applic-able to EDG testing. The log reco== ended by Section 3.a has been initiated. Data for tests since January 1, 1978, which were used in deter =ining the initial test frequency havt been entered in the log.
Future tests will be logged in accordance with the Guide.
The reporting reco=mendations of Section 3.b will be followed for all future failures.
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