ML17263A912
| ML17263A912 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 01/23/1995 |
| From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Andrea Johnson Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9502010125 | |
| Download: ML17263A912 (16) | |
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F'RI(DURITY ACCELERATED RIDS PROCESSING)
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION ÃBR:9502010125 DOC.DATE: 95/01/23 NOTARIZED: NO FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G
AUTH.NAME AUTHOR AFFILIATION MECREDY,R.C.
Rochester Gas
& Electric Corp.
RECIP.NAME RECIPIENT AFFILIATZON JOHNSON,A.R.
Project Directorate I-3 DOCKET 05000244 P
R
SUBJECT:
Forwards record of telcon which documents resolution of 7 remaining issues re emergency action levels.
DISTRIBUTION CODE: A045D COPIES RECEIVED: LTR Q ENCL SIZE:
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TITLE: OR Submittal:.Emergency Preparedness
- Plans, Impl ment'g Proce ures, C
0 NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).
05000244 R
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4ND ROCHESTER GASANDElECTRIC CORPORATION ~ 89 EASTAVENUE, ROCHESTER, N. Y 14649-0001 AREA CODE 716 5'-2700 ROBERT C. MECREDY Vice President Nvdeor Operotions January 23, 1995 U.S. Nuclear Regulatory Commission Document Control Desk Attn:
Allen R. Johnson Project Directorate I-3 Washington, D.C.
20555
Subject:
Record of Telecon December 16, 1994 Emergency Action Levels R.
E. Ginna Nuclear Power Plant Docket No. 50-244 Ref. (a): Letter from R.
C.
Mecredy (RG&E) to Allen R.
Johnson (NRC),
Emergency Action Levels
Response
to Request to Additional Xnformation, dated November 7, 1994
Dear Mr. Boynton:
Rochester Gas and Electric transmitted a response for additional information regarding emergency action levels by reference (a).
Subsequently, on 12/14/94 RG&E received a request to clarify 7
remaining issues.
A tele-conference was held on 12/16/94 between RG&E and NRC in order to clarify and provide final resolution to these open issues to support the issuance of the NRC safety evaluation report on the emergency action levels.
Enclosed is a
record of telecon which documents, the resolution of the 7 remaining issues.
Very truly yours, Robert C. Mecredy GAH4362 Oa Altos)
'st502010125 950123 PDR ADOCK 0~000 F
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xc:
Mr.. Allen R. Johnson (Mail Stop 14D1)
Project Directorate I-3
'ashington, D.C.
20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 U. S, Nuclear Regulatory Commission Attn:
Mr. Scott Boynton One White Flint North
( Mail Stop 9H15) 11155 Rockville Pike Rockville, MD 20852 Ginna Senior Resident Inspector
ENCLOSURE 1
The following responses are to the "Remaining Issues On R.E. Ginna Emergency Action Levels" telecopied to Mr. P. Polfleit on 12/14/94 and discussed in the telephone conversation between Mr. S. Boynton of the NRC, Mr. P. Polfleit of RGRE and Mr. K.
Walker of OSSI on 12/16/94.
Page 9, para.
1 Of RAI response The justification for originally not including the combination of >300pCi/cc DEI and >46 gpm RCS leakage was that this combination was covered by the containment rad monitor EAL of 100 R/hr. However, this philosophy is contrary to the philosophy in NUMARC/NESP-007 which suggests the use of multiple redundant indicators for entry into an emergency class.
Response to issue ¹1 "Per telephone discussions with S. Boynton, this issue is resolved and no futher information is required".
Page 18, last para. of RAI response Provide additional information on how evaluation and qualification of an RCS leak discharging outside containment is different from the evaluation and quantification of an RCS leak discharging inside containment (i.e. would an RCS leak outside containment affect a different response in the volume control system than one inside containment?).
Response to issue ¹2 The technical bases of EALs 4.1.3 and 4.1.6 have been revised to state the following:
"Inabilityto isolate any primary system discharging outside containment" is intended to address other primary systems, either direct or indirect, which the inability to isolate indicate loss of both RCS and containment.
No leakage threshold is specified since leaks outside containment, particularly under dynamic conditions, are difficultto quantify and may manifest themselves with diverse symptoms.
Symptoms of a primary system discharging outside containment may be indicated via mass balance, decreasing RCS inventory without corresponding containment response, or area temperatures and
'adiation levels outside containment.
It is for this reason that Shift Supervisor/Emergency Coordinator judgement is intended to be used in evaluating this criteria. However, it is intended that the magnitude of the leak associated with this EAL be consistent with the RCS barrier loss threshold of 46 gpm or greater.
Page 21, para. 3 of RAI response (For Information Only) During the development and endorsement of NUMARC/NESP-007, it was agreed that a loss of subcooling was a more
1
'A A
gL
'ppropriate indicator for loss of the RCS barrier for PWRs.
Thus, vessel water level was not included in the NUMARC/NESP-007 fission product barrier matrix as-an indicator for loss or potential loss of the RCS barrier.
In fact, a loss of subcooling, due to an RCS breach, would occur prior to reactor vessel water reaching the top of active fuel. The Site Area Emergency threshold would be reached no later than the time RVLIS indicates that fuel is being uncovered.
In regards to NUMARC/NESP-007 IC SS5, subcooling is not a valid indicator of RCS breach.
Also, due to the lack of a pressurized system and low driving head in the applicable modes for this IC, a loss of water level that leads to the imminent uncovery of fuel is indicative of a substantial RCS leak appropriate indicator for the loss of the RCS and potential loss of the fuel clad for these conditions.
Response to issue ¹3 "Per telephone discussions with S. Boynton, this issue is resolved and no futher information is required".
Page 26, para. 3 of RAI response The response states that the containment radiation monitor values were derived from assumptions in Regulatory Guide 1.25.
However, Regulatory Guide 1.25 is used for evaluating the dose consequences of a fuel handling accident, an accident that is only applicable in the refueling mode and one that is inconsistent with the use of this EAL. Provide additional information the exact fission product inventories that are assumed to be in containment from a 20% gap release at Ginna.
Response to issue ¹4 Refer to Attachment 1 for a table indicating the fission product inventories utilized for a 100% gap activity release used in the study from which the containment high range radiation monitor response curves were derived. The values were the result of the study performed for RG8 E by Technology for Energy Corporation (TEC) of Knoxville, Tenn. The study was titled "Modification Of The LEI Calculation Of Relationship Between In-Containment High Range Radiation Monitors And Releases Of Radioactive Material To The Containment Following An Accident." The purpose of this study was to develop values for the containment radiation monitors after a loss of coolant accident (LOCA). These inventories were based on the isotopic fractions specified in RG 1.25.
That is, an assumption that the gap activity seen by the containment radiation detectors consists of 10% of all noble gases other that Kr-85,30% of Kr-85, and 10% of the radioactive iodine at the time of the accident.
The isotopic mix was then dispersed inside the containment vessel and dose rates were calculated using the QADMOD-G computer code.
The dose rates were then charted for time 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after shutdown.
The value selected was based on 20% of the R-29 monitor response curve.
5.
Page 32, para. 3 of RAI response Additional information does not support the deviation from the guidance.
The failure of these systems does pose a significant threat to plant safety.
All of the Ginna EALs associated with failure of the reactor protection system should be revised to be consistent with the guidance in NUMARC/NESP-007.
With regards to the response's reference to NUMARC QRA General question
¹9, it is true that an emergency need not be declared if it is determined that an emergency condition no longer exists prior to the 15 minute offsite notification.
As stated in the QEA, these events should be classified and the classification and supporting event information reported to the NRC within one hour.
However, no information has been provided that adequately supports the argument that an emergency would no longer exist when the manual scram is successful.
It is expected that an emergency willbe declared for failure of the automatic reactor protection system to complete a scram and that the emergency continues to exist until it can be verified that the RPS failure resulted in no adverse consequences.
Response to issue ¹5 EAL 1.1.1 and EAL 1.1.2 have been revised to be consistent with NUIVIARC/NESP-007 and their technical bases revised (refer to the attached revised technical bases Attachment 2).
6.
Page 37, response to RAI ¹28 The response does not adequately explain why the PPCS and SAS were not explicitly referenced in the EAL.
Response to issue ¹6 "Per telephone discussions with S. Boynton, this issue is resolved and no futher information is required".
7.
Table 5.2, Dose Projection/Env. Measurement Classification Thresholds.
The terminology and units utilized in Table 5.2 are inconsistent with 10CFR Part 20:
Total Effective Dose Equivalent and Committed Dose Equivalent to the Thyroid are both quantified in rem, not rad as suggested by Table 5.2.
The use of TEDE rate and CDE Thyroid rate are inconsistent with 10 CFR Part 20 and are not defined as protective action guides in accordance with EPA-400.
The issue here is that a TEDE rate or CDE Thyroid rate can not be readily measured in the environment.
Also, when a dose assessment model generates TEDE and CDE Thyroid information, the quantity of interest is the total integrated dose over the entire release and exposure.
A rate of committed dose vs. exposure can't be evaluated against the protective action guidelines.
Response to issue ¹7 The terminology in Table 5.2 has been revised as follows:
Table 5.2 Dose Projection/Env. Measurement Classification Thresholds TEDE CDE Thyroid External exposure rate Thyroid exposure rate (for 1hr of inhalation)
ALERT 10 mRem N/A 10 mRem/hr N/A SAE 100 mRem 500 mRern 100 mRem/hr 500 mRem/hr GE 1000 mRem 5000 mRem 1000 mRem/hr 5000 mRem/hr
Attachment 1
Containment Activity(pCI/cc) for a 100% Release of the Gap Activity(RG 1.25)
ISOTOPE 0.0 0.5 (TIME hours) 2.0 8.0 Kr-85m Kr-85 Kr-87 Kr-88 Xe-131m Xe-133m Xe-1 33 Xe-1 35m Xe-135 I-1 31 l-1 32 I-1 33 1-1 34 1-135 Rb-88 TOTAL:
3.64+2*
4.26+0 6.92+1 1.06+2 0.00+0 7.28+0 2.88+2 0.00+0 5.46+1 8.00+1 1.09+2 1.46+2 1.68+2 1.31+2 0.00+0 1.53+3 3.36+2 1.01+1 5.28+1 9.36+1 8.54-4 7.29+0 2.90+2 9.88-1 5.81+1 7.99+1 9.37+1 1.44+2 1.13+2 1.24+2 7.21+0 1.41+3 2.66+2 2.60+1 2.35+1 6.45+1 3.40-3 7.32+0 2.94+2 1.16+0 6.88+1 7.94+1 5.95+1 1.37+2 3.45+1 1.06+2 7.64+0 1.17+3 1.03+2 5.92+1 9.20-1 1.45+1 1.34-2 7.36+0 3.09+2 6.26-1 8.30+1 7.78+1 9.70+0 1.12+2 2.98-1 6.68+1 1.73+0 8.35+2
- Note 3.64+2 denotes 3.64 x 10 '
1.0 CSFST Status 1.1 Subcriticality 1.1.1 Alert Any failure of an automatic trip signal to reduce power range <5% AND Manual trip is successful NUMARC IF:
Failure of Reactor Protection system instrumentation to complete or initiate an automatic reactor trip once a Reactor Protection system setpoint has been exceeded and manual trip was successful while in power operations or hot standby.
FPB Loss/potential loss N/A Mode Applicability:
Power Operations, Hot Shutdown Basis:
A manual trip is any set of actions by the reactor operator(s) at the reactor control console which causes "control rods to be rapidly inserted into the core and brings power below that percent power associated with the ability of the safety systems to remove heat and continue to decrease."
This condition indicates failure of the automatic protection systems to trip the reactor.
Failure of the manual trip to the extent requiring emergency operation requires declaration of a Site Area Emergency.
In determining whether to declare an emergency based on this EAL the following guidance is provided by NUMARC:
Regarding the occurrence of an event in which EAL is reached with no adverse consequences:
"Ifan emergency condition no longer exists, there is no reason to declare an emergency.
The NRC shall be notified after discovery within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, meeting 10CFR50.72 reporting criteria. State and local authorities should also be
notified as soon as practical, or in accordance with arrangements made in advance."
PEG
Reference:
SA2.1 Basis Reference(s) 1.
CSFST F-0.1, Subcriticality 2.
"Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 2 - Questions and Answers, June 1993
~
4 4
~
1.0 CSFST Status 1 ~ 1 Subcriticality 1.1.2 Site Area Emergency RED path in F-0.1 SUBCRITICALITYAND Emergency operation is required NUMARC IC:
Failure of Reactor Protection system instrumentation to complete or initiate an automatic reactor trip once a Reactor Protection system setpoint has been exceeded and manual scram trip was not successful.
FPB loss/potential loss N/A Mode Applicability:
Power Operations, Hot Shutdown Basis:
CSFST Subcriticality, RED path is entered based on failure of power range indication to decrease below 5% following a reactor trip. This portion of the EAL addresses any manual trip or automatic trip signal followed by a manual trip which fails to shut down the reactor to an extent that the reactor is producing more heat load for which the safety systems were designed.
A manual trip is any set of actions by the reactor operator(s) at the reactor control console which causes "control rods to be rapidly inserted into the core and brings power below that percent power associated with the ability of the safety systems to remove heat and continue to decrease."
This condition indicates failure of both the automatic and manual protection systems to trip the reactor to an extent that emergency operation is required.
The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat poses a direct threat to fuel clad and RCS integrity and thus warrants declaration of a Site Area Emergency.
~
A
PEG,Reterence:
SS2.1 SS4.1 Basis Reference(s):
2.
3.
4 CSFST F-0.1, Subcriticality FR-S.1, Response to Reactor Restart/ATWS FR-S.2, Response to Loss of Core Shutdown "Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 2 - Questions and Answers, June 1993